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Sample records for calcined high-level waste

  1. Process Design Concepts for Stabilization of High Level Waste Calcine

    Energy Technology Data Exchange (ETDEWEB)

    T. R. Thomas; A. K. Herbst

    2005-06-01

    The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels

  2. Spray Calciner/In-Can Melter high-level waste solidification technical manual

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E. (ed.)

    1980-09-01

    This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application.

  3. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory‘s Bench -Scale Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    Vince Maio

    2011-08-01

    This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine

  4. Summary of Waste Calcination at INTEC

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, Barry Henry; Newby, Bill Joe

    2000-10-01

    Fluidized-bed calcination at the Idaho Nuclear Technologies and Engineering Center (INTEC, formally called the Idaho Chemical Processing Plant) has been used to solidify acidic metal nitrate fuel reprocessing and incidental wastes wastes since 1961. A summary of waste calcination in full-scale and pilot plant calciners has been compiled for future reference. It contains feed compositions and operating conditions for all the processing campaigns for the original Waste Calcining Facility (WCF), the New Waste Calcining Facility (NWCF) started up in 1982, and numerous small scale pilot plant tests for various feed types. This summary provides a historical record of calcination at INTEC, and will be useful for evaluating calcinability of future wastes.

  5. High-Level Radioactive Waste.

    Science.gov (United States)

    Hayden, Howard C.

    1995-01-01

    Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

  6. Process for solidifying high-level nuclear waste

    Science.gov (United States)

    Ross, Wayne A.

    1978-01-01

    The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

  7. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    1999-06-01

    A potential option in the program for long-term management of high-level wastes at the Idaho Nuclear Technology and Engineering Center (INTEC), at the Idaho National Engineering and Environmental Laboratory, calls for retrieving calcine waste and converting it to a more stable and less dispersible form. An inventory of calcine produced during the period December 1963 to May 1999 has been prepared based on calciner run, solids storage facilities operating, and miscellaneous operational information, which gives the range of chemical compositions of calcine waste stored at INTEC. Information researched includes calciner startup data, waste solution analyses and volumes calcined, calciner operating schedules, solids storage bin capacities, calcine storage bin distributor systems, and solids storage bin design and temperature monitoring records. Unique information on calcine solids storage facilities design of potential interest to remote retrieval operators is given.

  8. Optimizing High Level Waste Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Dirk Gombert

    2005-09-01

    If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being

  9. Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Vienna, John D.; Peeler, David K.; Reamer, I. A.

    2001-03-30

    This report documents the results of glass formulation efforts for Idaho National Engineering and Environmental Laboratory (INEEL) high level waste (HWL) calcine. Two waste compositions were used during testing. Testing started by using the Run 78 calcine composition and switched to simulated Blend calcine composition when it became available. The goal of the glass formulation efforts was to develop a frit composition that will accept higher waste loading that satisfies the glass processing and product acceptance constraints. 1. Melting temperature of 1125 ? 25?C 2. Viscosity between 2 and 10 Pa?s at the melting temperature 3. Liquidus temperature at least 100?C below the melting temperature 4. Normalized release of B, Li and Na each below 1 g/m2 (per ASTM C 1285-97) Glass formulation efforts tested several frit compositions with variable waste loadings of Run 78 calcine waste simulant. Frit 107 was selected as the primary candidate for processing since it met all process and performance criteria up to 45 mass% waste loading. When the simulated Blend calcine waste composition became available Frits 107 and 108 compositions were retested and again Frit 107 remained the primary candidate. However, both frits suffered a decrease in waste loading when switching from the Run 78 calcine to simulated Blend calcine waste composition. This was due to increase concentrations of both F and Al2O3 along with a decrease in CaO and Na2O in the simulate Blend calcine waste all of which have strong impacts on the glass properties that limit waste loading of this type of waste.

  10. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

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    CERTA, P.J.

    2006-02-22

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

  11. Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Barry Scheetz; Johnson Olanrewaju

    2001-10-15

    The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

  12. HWMA closure plan for the Waste Calcining Facility at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-05-01

    The Waste Calcining Facility (WCF) calcined and evaporated aqueous wastes generated from the reprocessing of spent nuclear fuel. The calciner operated from 1963 to 1981, primarily processing high level waste from the first cycle of spent fuel extraction. Following the calciner shutdown the evaporator system concentrated high activity aqueous waste from 1983 until 1987. In 1988, US Department of Energy Idaho Operations Office (DOE-ID) requested interim status for the evaporator system, in anticipation of future use of the evaporator system. The evaporator system has not been operated since it received interim status. At the present time, DOE-ID is completing construction on a new evaporator at the New Waste Calcining Facility (NWCF) and the evaporator at the WCF is not needed. The decision to not use the WCF evaporator requires Lockheed Idaho Technologies Company (LITCO) and DOE-ID to close these units. After a detailed evaluation of closure options, LITCO and DOE-ID have determined the safest option is to fill the voids (grout the vessels, cells and waste pile) and close the WCF to meet the requirements applicable to landfills. The WCF will be covered with a concrete cap that will meet the closure standards. In addition, it was decided to apply these closure standards to the calcining system since it is contained within the WCF building. The paper describes the site, waste inventory, closure activities, and post-closure care plans.

  13. High-level waste qualification: Managing uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Pulsipher, B.A.

    1993-09-01

    A vitrification facility is being developed by the U.S. Department of Energy (DOE) at the West Valley Demonstration Plant (WVDP) near Buffalo, New York, where approximately 300 canisters of high-level nuclear waste glass will be produced. To assure that the produced waste form is acceptable, uncertainty must be managed. Statistical issues arise due to sampling, waste variations, processing uncertainties, and analytical variations. This paper presents elements of a strategy to characterize and manage the uncertainties associated with demonstrating that an acceptable waste form product is achieved. Specific examples are provided within the context of statistical work performed by Pacific Northwest Laboratory (PNL).

  14. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

    Perez Jr, Joseph M; Bickford, Dennis F; Day, Delbert E; Kim, Dong-Sang; Lambert, Steven L; Marra, Sharon L; Peeler, David K; Strachan, Denis M; Triplett, Mark B; Vienna, John D; Wittman, Richard S

    2001-07-13

    At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

  15. High-level radioactive wastes. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    McLaren, L.H. (ed.)

    1984-09-01

    This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

  16. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Staiger, Merle Daniel; M. C. Swenson

    2005-01-01

    This report documents an inventory of calcined waste produced at the Idaho Nuclear Technology and Engineering Center during the period from December 1963 to May 2000. The report was prepared based on calciner runs, operation of the calcined solids storage facilities, and miscellaneous operational information that establishes the range of chemical compositions of calcined waste stored at Idaho Nuclear Technology and Engineering Center. The report will be used to support obtaining permits for the calcined solids storage facilities, possible treatment of the calcined waste at the Idaho National Engineering and Environmental Laboratory, and to ship the waste to an off-site facility including a geologic repository. The information in this report was compiled from calciner operating data, waste solution analyses and volumes calcined, calciner operating schedules, calcine temperature monitoring records, and facility design of the calcined solids storage facilities. A compact disk copy of this report is provided to facilitate future data manipulations and analysis.

  17. Screening Level Risk Assessment for the New Waste Calcining Facility

    Energy Technology Data Exchange (ETDEWEB)

    M. L. Abbott; K. N. Keck; R. E. Schindler; R. L. VanHorn; N. L. Hampton; M. B. Heiser

    1999-05-01

    This screening level risk assessment evaluates potential adverse human health and ecological impacts resulting from continued operations of the calciner at the New Waste Calcining Facility (NWCF) at the Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Engineering and Environmental Laboratory (INEEL). The assessment was conducted in accordance with the Environmental Protection Agency (EPA) report, Guidance for Performing Screening Level Risk Analyses at Combustion Facilities Burning Hazardous Waste. This screening guidance is intended to give a conservative estimate of the potential risks to determine whether a more refined assessment is warranted. The NWCF uses a fluidized-bed combustor to solidify (calcine) liquid radioactive mixed waste from the INTEC Tank Farm facility. Calciner off volatilized metal species, trace organic compounds, and low-levels of radionuclides. Conservative stack emission rates were calculated based on maximum waste solution feed samples, conservative assumptions for off gas partitioning of metals and organics, stack gas sampling for mercury, and conservative measurements of contaminant removal (decontamination factors) in the off gas treatment system. Stack emissions were modeled using the ISC3 air dispersion model to predict maximum particulate and vapor air concentrations and ground deposition rates. Results demonstrate that NWCF emissions calculated from best-available process knowledge would result in maximum onsite and offsite health and ecological impacts that are less then EPA-established criteria for operation of a combustion facility.

  18. Intergenerational ethics of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Kunihiko [Nagoya Univ., Graduate School of Engineering, Nagoya, Aichi (Japan); Nasu, Akiko; Maruyama, Yoshihiro [Shibaura Inst. of Tech., Tokyo (Japan)

    2003-03-01

    The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

  19. 40 CFR 227.30 - High-level radioactive waste.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

  20. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Staiger, M. Daniel, Swenson, Michael C.

    2011-09-01

    This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

  1. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    2007-06-01

    This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

  2. High-Level waste process and product data annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Stegen, G.E.

    1996-02-13

    The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-level waste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references.

  3. Dissolution Kinetics of Alumina Calcine

    Energy Technology Data Exchange (ETDEWEB)

    Batcheller, Thomas Aquinas

    2001-09-01

    Dissolution kinetics of alumina type non-radioactive calcine was investigated as part of ongoing research that addresses permanent disposal of Idaho High Level Waste (HLW). Calcine waste was produced from the processing of nuclear fuel at the Idaho Nuclear Technology and Engineering Center (INTEC). Acidic radioactive raffinates were solidified at ~500°C in a fluidized bed reactor to form the dry granular calcine material. Several Waste Management alternatives for the calcine are presented in the Idaho High Level Waste Draft EIS. The Separations Alternative addresses the processing of the calcine so that the HLW is ready for removal to a national geological repository by the year 2035. Calcine dissolution is the key front-end unit operation for the separations alternative.

  4. Handbook of high-level radioactive waste transportation

    Energy Technology Data Exchange (ETDEWEB)

    Sattler, L.R.

    1992-10-01

    The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government`s system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government`s program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project.

  5. High-Level Waste System Process Interface Description

    Energy Technology Data Exchange (ETDEWEB)

    d' Entremont, P.D.

    1999-01-14

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

  6. FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M

    2008-05-09

    Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive High Level Waste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the high level radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

  7. Vitrification of high-level alumina nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Brotzman, J.R.

    1979-01-01

    Borophosphate glass compositions have been developed for the vitrification of a high-alumina calcined defense waste. The effect of substituting SiO/sub 2/, P/sub 2/O/sub 5/ and CuO for B/sub 2/O/sub 3/ on the viscosity and leach resistance was measured. The effect of the alkali to borate ratio and the Li/sub 2/O:Na/sub 2/O ratio on the melt viscosity and leach resistance was also measured.

  8. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables.

  9. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.

    2010-09-07

    High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

  10. In vitro studies evaluating leaching of mercury from mine waste calcine using simulated human body fluids

    Science.gov (United States)

    Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

    2010-01-01

    In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almadén, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute primary ore or compounds formed during Hg retorting. Elevated leachate Hg concentrations were found during calcine leaching using a simulated gastric fluid (as much as 6200 μg of Hg leached/g sample). Elevated Hg concentrations were also found in calcine leachates using a simulated lung fluid (as much as 9200 μg of Hg leached/g), serum-based fluid (as much as 1600 μg of Hg leached/g), and water of pH 5 (as much as 880 μg of Hg leached/g). The leaching capacity of Hg is controlled by calcine mineralogy; thus, calcines containing soluble Hg compounds contain higher leachate Hg concentrations. Results indicate that ingestion or inhalation of Hg mine-waste calcine may lead to increased Hg concentrations in the human body, especially through the ingestion pathway.

  11. High Level Waste (HLW) Feed Process Control Strategy

    Energy Technology Data Exchange (ETDEWEB)

    STAEHR, T.W.

    2000-06-14

    The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-level waste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-level waste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system.

  12. Corrosion and failure processes in high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Mahidhara, R.K.; Elleman, T.S.; Murty, K.L. [North Carolina State Univ., Raleigh, NC (United States)

    1992-11-01

    A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-Level Wastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

  13. Chemical and mineralogical modifications of simplified radioactive waste calcine during heat treatment

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, A. [CEA, DEN, DTCD/SECM/LDMC Marcoule, BP 17171, F-30207 Bagnols-sur-Cèze cedex (France); Schuller, S., E-mail: sophie.schuller@cea.fr [CEA, DEN, DTCD/SECM/LDMC Marcoule, BP 17171, F-30207 Bagnols-sur-Cèze cedex (France); Toplis, M.J. [CNRS, Observatoire Midi Pyrénées, IRAP (UMR 5277), F-31400 Toulouse (France); Podor, R.; Ravaux, J.; Clavier, N.; Brau, H.P. [ICSM UMR 5257 CEA/CNRS/UMR/ENSCM Site de Marcoule, BP 17171, F-30207 Bagnols-sur-Cèze cedex (France); Charpentier, T. [UMR CEA/CNRS 3299, IRAMIS, SIS2M, LSDRM Saclay, F-91191 Gif-sur-Yvette cedex (France); Angeli, F. [CEA, DEN, DTCD/SECM/LCLT Marcoule, BP 17171, F-30207 Bagnols-sur-Cèze cedex (France); Leterrier, N. [CEA, DEN, DM2S/SFME/LSFT Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2014-05-01

    The microstructural and mineralogical changes associated with heating calcined mixtures of Al(NO{sub 3}){sub 3}⋅9H{sub 2}O–NaNO{sub 3} have been studied. This system is a simplified analogue of high-level radioactive waste calcine, one of the raw materials used in the vitrification process employed for waste management. The decomposition (dehydration and denitration) and formation of secondary crystalline phases have been studied by differential thermal and gravimetric analysis (DTA and TGA), and heat-treated products characterized by X-ray diffraction, Raman spectroscopy, Nuclear Magnetic Resonance (NMR) and Transmission Electron Microscopy (TEM). It is found that pure Al(NO{sub 3}){sub 3}⋅9H{sub 2}O transforms to amorphous Al{sub 2}O{sub 3} at a temperature of ∼180 °C, well below that of the calcination process (500 °C). This amorphous Al{sub 2}O{sub 3} is highly porous with a high specific surface area, but may in turn convert to denser γ-Al{sub 2}O{sub 3} and α-Al{sub 2}O{sub 3} with increasing temperature. On the other hand, pure NaNO{sub 3} remains stable up to ∼880 °C, despite a solid–liquid transition at ∼320 °C. For Al(NO{sub 3}){sub 3}⋅9H{sub 2}O–NaNO{sub 3} mixtures, the products of calcination at 500 °C are found to consist of very fine porous material containing Na, Al and O, in addition to a variable proportion of well-defined crystals consisting of Na, and O. Heating these mixtures to temperatures of up to 1000 °C shows that for the case 80% Al(NO{sub 3}){sub 3}⋅9H{sub 2}O −20% NaNO{sub 3} (weight%) a variety of crystalline sodium aluminates is formed (NaAlO{sub 2}, NaAl{sub 11}O{sub 17}, NaAl{sub 6}O{sub 9.5}), while for the 50–50 mixture, only NaAlO{sub 2} is found. In large amounts, addition of alumina thus leads to the formation of crystalline phases rich in Al{sub 2}O{sub 3} that are responsible for hardening the calcine as the temperature rises. The kinetics of nitrogen loss from NaNO{sub 3} are also found to be

  14. Site suitability criteria for solidified high level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Heckman, R.A.; Holdsworth, T.; Towse, D.F.

    1979-03-07

    Activities devoted to development of regulations, criteria, and standards for storage of solidified high-level radioactive wastes are reported. The work is summarized in sections on site suitability regulations, risk calculations, geological models, aquifer models, human usage model, climatology model, and repository characteristics. Proposed additional analytical work is also summarized. (JRD)

  15. High-level waste management technology program plan

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

  16. Spent Fuel and High-Level Radioactive Waste Transportation Report

    Energy Technology Data Exchange (ETDEWEB)

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  17. Storage of High Level Nuclear Waste in Germany

    Directory of Open Access Journals (Sweden)

    Dietmar P. F. Möller

    2007-01-01

    Full Text Available Nuclear energy is very often used to generate electricity. But first the energy must be released from atoms what can be done in two ways: nuclear fusion and nuclear fission. Nuclear power plants use nuclear fission to produce electrical energy. The electrical energy generated in nuclear power plants does not produce polluting combustion gases but a renewable energy, an important fact that could play a key role helping to reduce global greenhouse gas emissions and tackling global warming especially as the electricity energy demand rises in the years ahead. This could be assumed as an ideal win-win situation, but the reverse site of the medal is that the production of high-level nuclear waste outweighs this advantage. Hence the paper attempt to highlight the possible state-of-art concepts for the safe and sustaining storage of high-level nuclear waste in Germany.

  18. ICPP radioactive liquid and calcine waste technologies evaluation final report and recommendation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    Using a formalized Systems Engineering approach, the Latched Idaho Technologies Company developed and evaluated numerous alternatives for treating, immobilizing, and disposing of radioactive liquid and calcine wastes at the Idaho Chemical Processing Plant. Based on technical analysis data as of March, 1995, it is recommended that the Department of Energy consider a phased processing approach -- utilizing Radionuclide Partitioning for radioactive liquid and calcine waste treatment, FUETAP Grout for low-activity waste immobilization, and Glass (Vitrification) for high-activity waste immobilization -- as the preferred treatment and immobilization alternative.

  19. Mixing Processes in High-Level Waste Tanks - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, P.F.

    1999-05-24

    The mixing processes in large, complex enclosures using one-dimensional differential equations, with transport in free and wall jets is modeled using standard integral techniques. With this goal in mind, we have constructed a simple, computationally efficient numerical tool, the Berkeley Mechanistic Mixing Model, which can be used to predict the transient evolution of fuel and oxygen concentrations in DOE high-level waste tanks following loss of ventilation, and validate the model against a series of experiments.

  20. SIMULANT DEVELOPMENT FOR SAVANNAH RIVER SITE HIGH LEVEL WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Stone, M; Russell Eibling, R; David Koopman, D; Dan Lambert, D; Paul Burket, P

    2007-09-04

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. The HLW consists of insoluble metal hydroxides (primarily iron, aluminum, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The HLW is processed in large batches through DWPF; DWPF has recently completed processing Sludge Batch 3 (SB3) and is currently processing Sludge Batch 4 (SB4). The composition of metal species in SB4 is shown in Table 1 as a function of the ratio of a metal to iron. Simulants remove radioactive species and renormalize the remaining species. Supernate composition is shown in Table 2.

  1. Glass-ceramic waste forms for immobilization of the fluorinel-sodium, alumina, and zirconia calcines stored at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Vinjamuri, K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1994-12-31

    Glass-ceramics appear to be very good candidate waste forms for immobilization of the calcined high level solid wastes, fluorinel-sodium (Fl/Na), alumina and zirconia that are stored at the Idaho Chemical Processing Plant (ICPP). Candidate experimental glass-ceramics were synthesized at ICPP by hot isostatically pressing (HIPing) a mixture of precompacted pilot plant calcine and additives. The glass-ceramic waste forms for immobilization of the Fl/Na, alumina, and zirconia calcines consist of 70 wt% Fl/Na calcine, 23.9 wt% SiO{sub 2}, 5 wt% Ti, 1.1 wt% B{sub 2}O{sub 3}; 70 wt% alumina calcine, 30 wt% SiO{sub 2}; and 70 wt% zirconia calcine, 20.25 wt% SiO{sub 2}, 5 wt% Ti, 2.25 wt% Na{sub 2}O, 1.75 wt% B{sub 2}O{sub 3}, 0.75 wt% Li{sub 2}O, respectively. The characteristics of the waste forms including density, chemical durability, glass and crystalline phases, and the microstructure are investigated. The 14-day MCC-1 total mass loss rates and the normalized elemental leach rates for aluminum, boron, calcium, cadmium, chromium, cesium, potassium, silicon, sodium, strontium, titanium, and zirconium are all less than 1 g/m{sup 2}-day. The crystalline phases for the Fl/Na and zirconia waste forms include zirconia, zircon, calcium fluoride, and titanates. In addition, cadmium sulphide in Fl/Na, and cadmium metal in zirconia waste form were also identified. The crystalline phases for the alumina waste form are alpha, gamma, and delta alumina, cristobalite, albite, and mullite. Glass phase separation was not observed in alumina and zirconia waste forms. The observed glass phase separation in Fl/Na waste form appears to be chemically durable.

  2. Technology development at the Pacific Northwest Laboratory high-level waste management history

    Energy Technology Data Exchange (ETDEWEB)

    McElroy, J.L. [Geosafe Corp., Richland, WA (United States); Platt, A.M.

    1996-12-31

    During WWII and the post-WWII years, until the late 1950`s, plutonium production was Hanford`s primary mission. This mission produced an enormous legacy of wastes that have themselves become the new mission at Hanford. Waste management, as practiced at Hanford, during the defense production years was in many ways unique to Hanford, taking advantage of the dry climate, distance from the Columbia river and depth to the water table. Near-surface storage in tanks, ion exchange in seepage trenches and cribs, and near surface burial were the norm. Isolation of the wastes by the high and dry nature of the 200 Area plateau, where reprocessing and waste management took place, was one of the reasons Hanford had been selected for it`s nuclear mission. Thus, location was a significant aspect of the initial waste management program at Hanford. Treatment, other than simple chemical steps such as neutralization and ion exchange, had not been considered necessary to the mission and was therefore not developed. To support the development of commercial nuclear power and to provide improved means of handling nuclear wastes, new waste management programs were initiated in the 1950`s by the Atomic Energy Commission. The programs focused on high level waste. They included `spray calcination/vitrification` at Hanford Laboratories. Hanford Labs later became Pacific Northwest Laboratories (PNL) when Battelle Memorial Institute became the Operating Contractor in 1965. In 1996, it was renamed Pacific Northwest National Laboratory (PNNL). The purpose of this paper is to describe the HLW projects and programs that followed from this early HLW R&D at PNNL.

  3. 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

  4. High-level waste tank farm set point document

    Energy Technology Data Exchange (ETDEWEB)

    Anthony, J.A. III

    1995-01-15

    Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

  5. Review of High Level Waste Tanks Ultrasonic Inspection Data

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B

    2006-03-09

    A review of the data collected during ultrasonic inspection of the Type I high level waste tanks has been completed. The data was analyzed for relevance to the possibility of vapor space corrosion and liquid/air interface corrosion. The review of the Type I tank UT inspection data has confirmed that the vapor space general corrosion is not an unusually aggressive phenomena and correlates well with predicted corrosion rates for steel exposed to bulk solution. The corrosion rates are seen to decrease with time as expected. The review of the temperature data did not reveal any obvious correlations between high temperatures and the occurrences of leaks. The complex nature of temperature-humidity interaction, particularly with respect to vapor corrosion requires further understanding to infer any correlation. The review of the waste level data also did not reveal any obvious correlations.

  6. 4.5 Meter high level waste canister study

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R. B.

    1997-10-01

    The Tank Waste Remediation System (TWRS) Storage and Disposal Project has established the Immobilized High-Level Waste (IBLW) Storage Sub-Project to provide the capability to store Phase I and II BLW products generated by private vendors. A design/construction project, Project W-464, was established under the Sub-Project to provide the Phase I capability. Project W-464 will retrofit the Hanford Site Canister Storage Building (CSB) to accommodate the Phase I I-ILW products. Project W-464 conceptual design is currently being performed to interim store 3.0 m-long BLW stainless steel canisters with a 0.61 in diameter, DOE is considering using a 4.5 in canister of the same diameter to reduce permanent disposal costs. This study was performed to assess the impact of replacing the 3.0 in canister with the 4.5 in canister. The summary cost and schedule impacts are described.

  7. Calcined solids storage facility closure study

    Energy Technology Data Exchange (ETDEWEB)

    Dahlmeir, M.M.; Tuott, L.C.; Spaulding, B.C. [and others

    1998-02-01

    The disposal of radioactive wastes now stored at the Idaho National Engineering and Environmental Laboratory is currently mandated under a {open_quotes}Settlement Agreement{close_quotes} (or {open_quotes}Batt Agreement{close_quotes}) between the Department of Energy and the State of Idaho. Under this agreement, all high-level waste must be treated as necessary to meet the disposal criteria and disposed of or made road ready to ship from the INEEL by 2035. In order to comply with this agreement, all calcined waste produced in the New Waste Calcining Facility and stored in the Calcined Solids Facility must be treated and disposed of by 2035. Several treatment options for the calcined waste have been studied in support of the High-Level Waste Environmental Impact Statement. Two treatment methods studied, referred to as the TRU Waste Separations Options, involve the separation of the high-level waste (calcine) into TRU waste and low-level waste (Class A or Class C). Following treatment, the TRU waste would be sent to the Waste Isolation Pilot Plant (WIPP) for final storage. It has been proposed that the low-level waste be disposed of in the Tank Farm Facility and/or the Calcined Solids Storage Facility following Resource Conservation and Recovery Act closure. In order to use the seven Bin Sets making up the Calcined Solids Storage Facility as a low-level waste landfill, the facility must first be closed to Resource Conservation and Recovery Act (RCRA) standards. This study identifies and discusses two basic methods available to close the Calcined Solids Storage Facility under the RCRA - Risk-Based Clean Closure and Closure to Landfill Standards. In addition to the closure methods, the regulatory requirements and issues associated with turning the Calcined Solids Storage Facility into an NRC low-level waste landfill or filling the bin voids with clean grout are discussed.

  8. Deep borehole disposal of high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

    2009-07-01

    Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

  9. Heat Transfer Characteristics of Calcined Petroleum Coke in Waste Heat Recovery Process

    Directory of Open Access Journals (Sweden)

    Bin Zheng

    2016-01-01

    Full Text Available This paper reports the results of heat transfer characteristics of calcined petroleum coke in waste heat recovery process. The model of heat exchanger was set up. The model has been used to investigate the effects of porosity (0.58 to 0.79, equivalent heat conductivity coefficient (0.9 to 1.1, and equivalent specific heat (0.9 to 1.1. The calculated values of calcined petroleum coke temperature showed good agreement with the corresponding available experimental data. The temperature distribution of calcined petroleum coke, the calcined petroleum coke temperature at heat exchanger outlet, the average heat transfer coefficient, and the heat recovery efficiency were studied. It can also be used in deriving much needed data for heat exchanger designs when employed in industry.

  10. Defense High-Level Waste Leaching Mechanisms Program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E. (compiler)

    1984-08-01

    The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

  11. CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Burns, H.; Stefanko, D.

    2012-01-10

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by

  12. High level nuclear waste treatment in the Defense Waste Processing Facility: Overview and integrated flowsheet model

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.S.; Fowler, J.R.; Edwards, R.E. Jr.; Randall, C.T.

    1991-12-31

    Design and construction of the world`s largest vitrification facility for high level nuclear waste has been nearly completed at the US Department of Energy`s Savannah River Site. Equipment testing and calibration are currently being performed in preparation for the nonradioactive Chemical Runs in the late 1991. In 1993, the Defense Waste Processing Facility (DWPF) will begin producing 100 kg/hr of radioactive waste glass at 28 wt% waste oxide loading. This paper describes all phases of waste processing operations in DWPF and waste tank farms using the integrated flowsheet modeling approach. Particular emphases are given to recent developments in the DWPF processes and design.

  13. High Level Waste System Impacts from Acid Dissolution of Sludge

    Energy Technology Data Exchange (ETDEWEB)

    KETUSKY, EDWARD

    2006-04-20

    This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site High Level Waste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

  14. Transmutation of high-level radioactive waste - Perspectives

    CERN Document Server

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

  15. Control of high-level radioactive waste-glass melters

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter.

  16. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Richardson

    2003-03-19

    In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

  17. Stability of High-Level Radioactive Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, T.M.

    2001-06-22

    High-level waste (HLW) glass compositions, processing schemes, limits on waste content, and corrosion/dissolution release models are dependent on an accurate knowledge of melting temperatures and thermochemical values. Unfortunately, existing models for predicting these temperatures are empirically-based, depending on extrapolations of experimental information. In addition, present models of leaching behavior of glass waste forms use simplistic assumptions or experimentally measured values obtained under non-realistic conditions. There is thus a critical need for both more accurate and more widely applicable models for HLW glass behavior, which this project addressed. Significant progress was made in this project on modeling HLW glass. Borosilicate glass was accurately represented along with the additional important components that contain iron, lithium, potassium, magnesium, and calcium. The formation of crystalline inclusions in the glass, an issue in Hanford HLW formulations, was modeled and shown to be predictive. Thus the results of this work have already demonstrated practical benefits with the ability to map compositional regions where crystalline material forms, and therefore avoid that detrimental effect. With regard to a fundamental understanding, added insights on the behavior of the components of glass have been obtained, including the potential formation of molecular clusters. The EMSP project had very significant effects beyond the confines of Environmental Management. The models developed for glass have been used to solve a very costly problem in the corrosion of refractories for glass production. The effort resulted in another laboratory, Sandia National Laboratories-Livermore, to become conversant in the techniques and to apply those through a DOE Office of Industrial Technologies project joint with PPG Industries. The glass industry as a whole is now cognizant of these capabilities, and there is a Glass Manufacturer's Research Institute

  18. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  19. Alternative calcination development status report

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, R.D.

    1997-12-01

    The Programmatic Spent Nuclear Fuel and (INEEL) Environmental Restoration and Waste Management Programs Environmental Impact Statement Record of Decision, dated June 1, 1995, specifies that high-level waste stored in the underground tanks at the ICPP continue to be calcined while other options to treat the waste are studied. Therefore, the High-Level Waste Program has funded a program to develop new flowsheets to increase the liquid waste processing rate. Simultaneously, a radionuclide separation process, as well as other options, are also being developed, which will be compared to the calcination treatment option. Two alternatives emerged as viable candidates; (1) elevated temperature calcination (also referred to as high temperature calcination), and (2) sugar-additive calcination. Both alternatives were determined to be viable through testing performed in a lab-scale calcination mockup. Subsequently, 10-cm Calciner Pilot Plant scoping tests were successfully completed for both flowsheets. The results were compared to the standard 500 C, high-ANN flow sheet (baseline flowsheet). The product and effluent streams were characterized to help elucidate the process chemistry and to investigate potential environmental permitting issues. Several supplementary tests were conducted to gain a better understanding of fine-particles generation, calcine hydration, scrub foaming, feed makeup procedures, sugar/organic elimination, and safety-related issues. Many of the experiments are only considered to be scoping tests, and follow-up experiments will be required to establish a more definitive understanding of the flowsheets. However, the combined results support the general conclusion that flowsheet improvements for the NWCF are technically viable.

  20. Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Steven E.

    2013-11-11

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

  1. Parametric Analyses of Heat Removal from High Level Waste Tanks

    Energy Technology Data Exchange (ETDEWEB)

    TRUITT, J.B.

    2000-06-05

    The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-02

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

  3. Tank waste remediation system high-level waste feed processability assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, S.L. [Westinghouse Hanford Co., Richland, WA (United States); Kim, D.S. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01

    This study evaluates the effect of feed composition on the performance of the high-level vitrification process. It is assumed in this study that the tank wastes are retrieved and blended by tank farms, producing 12 different blends from the single-shell tank farms, two blends of double-shell tank waste, and a separately defined all-tank blend. This blending scenario was chosen only for evaluating the impact of composition on the volume of high- level waste glass produced. Special glass compositions were formulated for each waste blend based on glass property models and the properties of similar glasses. These glasses were formulated to meet the applicable viscosity, electrical conductivity, and liquidus temperature constraints for the identified candidate melters. Candidate melters in this study include the low-temperature stirred melter, which operates at 1050{degrees}C; the reference Hanford Waste Vitrification Plant liquid-fed ceramic melter, which operates at 1150{degrees}C; and the high-temperature, joule-heated melter and the cold-crucible melter, which operate over a temperature range of 1150{degrees}C to 1400{degrees}C. In the most conservative case, it is estimated that 61,000 MT of glass will be produced if the Site`s high-level wastes are retrieved by tank farms and processed in the reference joule-heated melter. If an all-tank blend was processed under the same conditions, the reference melter would produce 21,250 MT of glass. If cross-tank blending were used, it is anticipated that $2.0 billion could be saved in repository disposal costs (based on an average disposal cost of $217,000 per canister) by blending the S, SX, B, and T Tank Farm wastes with other wastes prior to vitrification. General blending among all the tank farms is expected to produce great potential benefit.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  5. High-Level Radioactive Waste: Safe Storage and Ultimate Disposal.

    Science.gov (United States)

    Dukert, Joseph M.

    Described are problems and techniques for safe disposal of radioactive waste. Degrees of radioactivity, temporary storage, and long-term permanent storage are discussed. Included are diagrams of estimated waste volumes to the year 2000 and of an artist's conception of a permanent underground disposal facility. (SL)

  6. Inventory evaluation of high level radioactive vitrified waste

    Energy Technology Data Exchange (ETDEWEB)

    Ishihara, Yoshinao; Makino, Hitoshi; Ohi, Takao; Ishiguro, Katsuhiko [Japan Nuclear Cycle Development Inst., Waste Management and Fuel Cycle Research Center, Tokai Works, Tokai, Ibaraki (Japan); Miyahara, Kaname; Umeki, Hiroyuki [Japan Nuclear Cycle Development Inst., Geological Isolation Research Project, Tokai, Ibaraki (Japan); Akasaka, Hidenari; Fujihara, Hiroshi [Tokyo Electirc Power Company, Tokyo (Japan); Hashi, Yasuyuki [Computer Software Development Co., Ltd., Tokyo (Japan)

    1999-11-01

    Spent fuel removed from nuclear power plants in Japan is reprocessed at home and abroad (Japan Nuclear Fuel Limited [JNFL] reprocessing plant, Tokai Vitrification Facility [TVF] of JNC, COGEMA in France and BNFL in England) and then vitrified. The properties of vitrified waste change depending on the fuel type, fuel burn-up in the reactor, and operating conditions at the reprocessing plants. However, properties of vitrified waste, such as heat generation and nuclide inventories, are important information for the design study of repository and the performance assessment in the geological disposal system. For the objectives of repository design and safety assessment, it is necessary that the model-vitrified waste is determined from the four types of waste. In this study, for supporting the determination of the model vitrified waste in the Research and Development for the Geological Disposal of HLW in Japan (H12 Project), the calculation and comparison of inventories for the four types of waste were performed. As results, it is found that there are no significant differences in the properties (i.e., radioactivity, heat generation, hazard index and nuclide inventories per one package) of the four types of vitrified waste from JNFL, COGEMA, BNFL and TVF. (author)

  7. Demonstration of Caustic-Side Solvent Extraction with Savannah River Site High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Walker, D.D.

    2001-08-27

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet for the decontamination of high level waste using a 33-stage, 2-cm centrifugal contactor apparatus at the Savannah River Technology Center. This represents the first CSSX process demonstration using Savannah River Site (SRS) high level waste. Three tests lasting 6, 12, and 48 hours processed simulated average SRS waste, simulated Tank 37H/44F composite waste, and Tank 37H/44F high level waste, respectively.

  8. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of 2017 experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-11

    A full-scale, transparent mock-up of the Hanford Tank Waste Treatment and Immobilization Project High Level Waste glass melter riser and pour spout has been constructed to allow for testing with visual feedback of particle settling, accumulation, and resuspension when operating with a controlled fraction of crystals in the glass melt. Room temperature operation with silicone oil and magnetite particles simulating molten glass and spinel crystals, respectively, allows for direct observation of flow patterns and settling patterns. The fluid and particle mixture is recycled within the system for each test.

  9. Iron Phosphate Glass as Potential Waste Matrix for High-Level Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Fukui, T.; Ishinomori, T.; Endo, Y.; Sazarashi, M.; Ono, S.; Suzuki, K.

    2003-02-25

    Recently, Iron Phosphate Glass (IPG) is investigated as the alternative final waste form for High-Level Radioactive Waste (HLW) in U.S. This study is aimed to investigate feasibility of IPG to HLW arising from commercial reprocessing in Japan. In order to evaluate favorable preparation conditions, maximum waste loading and property of IPG, the melting tests were carried. From the results of melting tests, the favorable preparation conditions was with matrix of Fe/P 0.43 (mole ratio in products) and melting at 1200{sup o} for 4h. The products of 10-20mass% waste loading of simulated HLW were glassy and had no crystal peaks, however the product of 30mass% waste loading showed some crystal peaks by XRD analysis. IPG and Borosilicate glass (BG) had about the same thermal properties. As a result, IPG had enough potential for high waste loading and the extremely good chemical durability for consideration as a waste form for Japanese HLW.

  10. Idaho National Engineering and Environmental Laboratory, Old Waste Calcining Facility, Scoville vicinity, Butte County, Idaho -- Photographs, written historical and descriptive data. Historical American engineering record

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    This report describes the history of the Old Waste Calcining Facility. It begins with introductory material on the Idaho National Engineering and Environmental Laboratory, the Materials Testing Reactor fuel cycle, and the Idaho Chemical Processing Plant. The report then describes management of the wastes from the processing plant in the following chapters: Converting liquid to solid wastes; Fluidized bed waste calcining process and the Waste Calcining Facility; Waste calcining campaigns; WCF gets a new source of heat; New Waste Calcining Facility; Last campaign; Deactivation and the RCRA cap; Significance/context of the old WCF. Appendices contain a photo key map for HAER photos, a vicinity map and neighborhood of the WCF, detailed description of the calcining process, and chronology of WCF campaigns.

  11. Strontium and Actinide Separations from High Level Nuclear Waste Solutions using Monosodium Titanate - Actual Waste Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T.B.; Barnes, M.J.; Hobbs,D.T.; Walker, D.D.; Fondeur, F.F.; Norato, M.A.; Pulmano, R.L.; Fink, S.D.

    2005-11-01

    Pretreatment processes at the Savannah River Site will separate {sup 90}Sr, alpha-emitting and radionuclides (i.e., actinides) and {sup 137}Cs prior to disposal of the high-level nuclear waste. Separation of {sup 90}Sr and alpha-emitting radionuclides occurs by ion exchange/adsorption using an inorganic material, monosodium titanate (MST). Previously reported testing with simulants indicates that the MST exhibits high selectivity for strontium and actinides in high ionic strength and strongly alkaline salt solutions. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from actual waste solutions. These tests evaluated the effects of ionic strength, mixing, elevated alpha activities, and multiple contacts of the waste with MST. Tests also provided confirmation that MST performs well at much larger laboratory scales (300-700 times larger) and exhibits little affinity for desorption of strontium and plutonium during washing.

  12. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    Energy Technology Data Exchange (ETDEWEB)

    Marzolf, A.; Folsom, M.

    2010-08-31

    This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame

  13. Characterization and Dissolution Kinetics Testing of Radioactive H-3 Calcine

    Energy Technology Data Exchange (ETDEWEB)

    Garn, Troy Gerry; Batcheller, Thomas Aquinas

    2002-09-01

    Characterization and dissolution kinetics testing were performed with Idaho radioactive H-3 calcine. Calcine dissolution is the key front-end unit operation for the Separations Alternative identified in the Idaho High Level Waste Draft EIS. The impact of the extent of dissolution on the feasibility of Separations must be clearly quantified.

  14. 75 FR 61228 - Board Meeting: Technical Lessons Gained From High-Level Nuclear Waste Disposal Efforts

    Science.gov (United States)

    2010-10-04

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR WASTE TECHNICAL REVIEW BOARD Board Meeting: Technical Lessons Gained From High-Level Nuclear Waste Disposal Efforts Pursuant to its authority under section 5051 of Public Law 100-203, Nuclear Waste Policy Amendments Act of...

  15. Transmutation of LWR high-level waste in LMRs

    Energy Technology Data Exchange (ETDEWEB)

    Dobbin, K.D.; Wootan, D.W.; Rawlins, J.A. (Westinghouse Hanford Co., Richland, WA (United States)); Jordheim, D.P.

    1991-11-01

    Past studies have shown that actinide waste discharged from light water reactors (LWRs) could be burned effectively in liquid-metal fast reactor systems. These systems were predominantly mixed plutonium/uranium oxide or metal breeder systems with small percentages of minor actinides. This study illustrates how LWR actinides and fission products could be burned in a minor actinide nitride-fueled advanced liquid-metal reactor (ALMR) that has special moderating assemblies to enhance fission product transmutation. For this study, the reactor was assumed to be only a burner, not a breeder. The model contained a preponderance of minor actinides with only sufficient plutonium to maintain 1-yr fuel cycles. No uranium was loaded because plutonium production was not desired.

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

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); 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).

  17. Development of polyphase ceramics for the immobilization of high-level Defense nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, P.E.D.; Harker, A.B.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

    1983-02-25

    The report contains two major sections: Section I - An Improved Polyphase Ceramic for High-Level Defense Nucleation Waste reports the work conducted on titanium-silica based ceramics for immobilizing Savannah River Plant waste. Section II - Formulation and Processing of Alumina Based Ceramic Nuclear Waste Forms describes the work conducted on developing a generic alumina and alumina-silica based ceramic waste form capable of immobilizing any nuclear waste with a high aluminum content. Such wastes include the Savannah River Plant wastes, Hanford neutralized purex wastes, and Hanford N-Reactor acid wastes. The design approach and process technology in the two reports demonstrate how the generic high waste loaded ceramic form can be applied to a broad range of nuclear waste compositions. The individual sections are abstracted and indexed separately.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-04

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

  19. Demonstration of Small Tank Tetraphenylborate Precipitation Process Using Savannah River Site High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T.B.

    2001-09-10

    This report details the experimental effort to demonstrate the continuous precipitation of cesium from Savannah River Site High Level Waste using sodium tetraphenylborate. In addition, the experiments examined the removal of strontium and various actinides through addition of monosodium titanate.

  20. Research and development on geological disposal of high-level radioactive waste; First progress report

    OpenAIRE

    1992-01-01

    The "first progress report of research and development ongeological disposal of high level radioactive waste", h3 in short, isintended for the japanese authorities. In accordance with the "overallprogram for high level radioactive waste management" set forth byatomic energy commission, h3 is designed to clarify the current status ofthe research and development work performed by power reactor and nuclearfuel development corporation(pnc) up to the year 1991. H3 presents the updated knowledge on...

  1. H-3 Summary report research and development on geolgical disposal of high-level radioactive waste

    OpenAIRE

    1992-01-01

    The "First progress report of research and development ongeological disposal of high level radioactive waste",H3 in short,is intended for the Japanese authorities. In accordance with the "Overall program for high level radioactive waste management" set forth by atomic energy commission, H3 is designed to clarify the current status of the research and development work performed by power reactor and nuclear fuel development corporation (PNC) up to the year 1991. H3 presents the updated knowledg...

  2. Chem I Supplement. Chemistry Related to Isolation of High-Level Nuclear Waste.

    Science.gov (United States)

    Hoffman, Darleane C.; Choppin, Gregory R.

    1986-01-01

    Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

  3. West Valley demonstration project: alternative processes for solidifying the high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

    1981-10-01

    In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

  4. Biodiesel production from palm oil using calcined waste animal bone as catalyst.

    Science.gov (United States)

    Obadiah, Asir; Swaroopa, Gnanadurai Ajji; Kumar, Samuel Vasanth; Jeganathan, Kenthorai Raman; Ramasubbu, Alagunambi

    2012-07-01

    Waste animal bones was employed as a cost effective catalyst for the transesterification of palm oil. The catalyst was calcined at different temperatures to transform the calcium phosphate in the bones to hydroxyapatite and 800 °C was found to give the best yield of biodiesel. The catalyst was characterized by scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive spectrometry (EDS) and Fourier transform infrared spectrometry (FT-IR). Under the optimal reaction conditions of 20 wt.% of catalyst, 1:18 oil to methanol molar ratio, 200 rpm of stirring of reactants and at a temperature of 65 °C, the methyl ester conversion was 96.78% and it was achieved in 4h. The catalyst performed equally well as the laboratory-grade CaO. Animal bone is therefore a useful raw material for the production of a cheap catalyst for transesterification. Copyright © 2012 Elsevier Ltd. All rights reserved.

  5. Composite quarterly technical report long-term high-level-waste technology, October-December 1981

    Energy Technology Data Exchange (ETDEWEB)

    Cornman, W.R. (comp.)

    1982-06-01

    This document summarizes work performed at participating sites on the immobilization of high-level wastes from the chemical reprocessing of reactor fuels. The plan is to develop waste form alternatives for each of the three DOE sites (SRP, ICPP, and Hanford). Progress is reported in the following areas: waste preparation; fixation in glass, concrete, tailored ceramics, and coated particles; process and equipment development; and final handling. 12 figures, 19 tables. (DLC)

  6. Mercury Phase II Study - Mercury Behavior across the High-Level Waste Evaporator System

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jackson, D. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shah, H. B. [Savannah River Remediation, LLC., Aiken, SC (United States); Jain, V. [Savannah River Remediation, LLC., Aiken, SC (United States); Occhipinti, J. E. [Savannah River Remediation, LLC., Aiken, SC (United States); Wilmarth, W. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-06-17

    The Mercury Program team’s effort continues to develop more fundamental information concerning mercury behavior across the liquid waste facilities and unit operations. Previously, the team examined the mercury chemistry across salt processing, including the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU), and the Defense Waste Processing Facility (DWPF) flowsheets. This report documents the data and understanding of mercury across the high level waste 2H and 3H evaporator systems.

  7. Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    2000-02-17

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

  8. Environmental assessment: Closure of the Waste Calcining Facility (CPP-633), Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The U.S. Department of Energy (DOE) proposes to close the Waste Calcining Facility (WCF). The WCF is a surplus DOE facility located at the Idaho Chemical Processing Plant (ICPP) on the Idaho National Engineering Laboratory (INEL). Six facility components in the WCF have been identified as Resource Conservation and Recovery Ace (RCRA)-units in the INEL RCRA Part A application. The WCF is an interim status facility. Consequently, the proposed WCF closure must comply with Idaho Rules and Standards for Hazardous Waste contained in the Idaho Administrative Procedures Act (IDAPA) Section 16.01.05. These state regulations, in addition to prescribing other requirements, incorporate by reference the federal regulations, found at 40 CFR Part 265, that prescribe the requirements for facilities granted interim status pursuant to the RCRA. The purpose of the proposed action is to reduce the risk of radioactive exposure and release of hazardous constituents and eliminate the need for extensive long-term surveillance and maintenance. DOE has determined that the closure is needed to reduce potential risks to human health and the environment, and to comply with the Idaho Hazardous Waste Management Act (HWMA) requirements.

  9. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

  10. Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

    Science.gov (United States)

    Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

    2012-06-05

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

  11. Development of a test system for high level liquid waste partitioning

    Directory of Open Access Journals (Sweden)

    Duan Wu H.

    2015-01-01

    Full Text Available The partitioning and transmutation strategy has increasingly attracted interest for the safe treatment and disposal of high level liquid waste, in which the partitioning of high level liquid waste is one of the critical technical issues. An improved total partitioning process, including a tri-alkylphosphine oxide process for the removal of actinides, a crown ether strontium extraction process for the removal of strontium, and a calixcrown ether cesium extraction process for the removal of cesium, has been developed to treat Chinese high level liquid waste. A test system containing 72-stage 10-mm-diam annular centrifugal contactors, a remote sampling system, a rotor speed acquisition-monitoring system, a feeding system, and a video camera-surveillance system was successfully developed to carry out the hot test for verifying the improved total partitioning process. The test system has been successfully used in a 160 hour hot test using genuine high level liquid waste. During the hot test, the test system was stable, which demonstrated it was reliable for the hot test of the high level liquid waste partitioning.

  12. Application of new technologies for characterization of Hanford Site high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Winters, W.I.

    1998-02-03

    To support remediation of Hanford Site high-level radioactive waste tanks, new chemical and physical measurement technologies must be developed and deployed. This is a major task of the Chemistry Analysis Technology Support (CATS) group of the Hanford Corporation. New measurement methods are required for efficient and economical resolution of tank waste safety, waste retrieval, and disposal issues. These development and deployment activities are performed in cooperation with Waste Management Federal Services of Hanford, Inc. This paper provides an overview of current analytical technologies in progress. The high-level waste at the Hanford Site is chemically complex because of the numerous processes used in past nuclear fuel reprocessing there, and a variety of technologies is required for effective characterization. Programmatic and laboratory operational needs drive the selection of new technologies for characterizing Hanford Site high-level waste, and these technologies are developed for deployment in laboratories, hot cells or in the field. New physical methods, such as the propagating reactive systems screening tool (PRSST) to measure the potential for self-propagating reactions in stored wastes, are being implemented. Technology for sampling and measuring gases trapped within the waste matrix is being used to evaluate flammability hazards associated with gas releases from stored wastes. Application of new inductively coupled plasma and laser ablation mass spectrometry systems at the Hanford Site`s 222-S Laboratory will be described. A Raman spectroscopy probe mounted in a cone penetrometer to measure oxyanions in wastes or soils will be described. The Hanford Site has used large volumes of organic complexants and acids in processing waste, and capillary zone electrophoresis (CZE) methods have been developed for determining several of the major organic components in complex waste tank matrices. The principles involved, system installation, and results from

  13. Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1) Introduction

    Energy Technology Data Exchange (ETDEWEB)

    R.A. Levich; J.S. Stuckless

    2006-09-25

    Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

  14. Determination of performance criteria for high-level solidified nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Heckman, R.A.; Holdsworth, T.

    1979-05-07

    To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste.

  15. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage.

  16. Phase I high-level waste pretreatment and feed staging plan

    Energy Technology Data Exchange (ETDEWEB)

    Manuel, A.F.

    1996-02-05

    This document provides the preliminary planning basis for the U.S. Department of Energy (DOE) to provide a sufficient quantity of high-level waste feed to the privatization contractor during Phase I. By this analysis of candidate high-level waste feed sources, the initial quantity of high-level waste feed totals more than twice the minimum feed requirements. The flexibility of the current infrastructure within tank farms provides a variety of methods to transfer the feed to the privatization contractor`s site location. The amount and type of pretreatment (sludge washing) necessary for the Phase I processing can be tailored to support the demonstration goals without having a significant impact on glass volume (i.e., either inhibited water or caustic leaching can be used).

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

    Science.gov (United States)

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

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

  18. Specifying the Concept of Future Generations for Addressing Issues Related to High-Level Radioactive Waste.

    Science.gov (United States)

    Kermisch, Celine

    2016-12-01

    The nuclear community frequently refers to the concept of "future generations" when discussing the management of high-level radioactive waste. However, this notion is generally not defined. In this context, we have to assume a wide definition of the concept of future generations, conceived as people who will live after the contemporary people are dead. This definition embraces thus each generation following ours, without any restriction in time. The aim of this paper is to show that, in the debate about nuclear waste, this broad notion should be further specified and to clarify the related implications for nuclear waste management policies. Therefore, we provide an ethical analysis of different management strategies for high-level waste in the light of two principles, protection of future generations-based on safety and security-and respect for their choice. This analysis shows that high-level waste management options have different ethical impacts across future generations, depending on whether the memory of the waste and its location is lost, or not. We suggest taking this distinction into account by introducing the notions of "close future generations" and "remote future generations", which has important implications on nuclear waste management policies insofar as it stresses that a retrievable disposal has fewer benefits than usually assumed.

  19. Research about the pozzolanic activity of waste materials from calcined clay

    Directory of Open Access Journals (Sweden)

    Sánchez de Rojas, M. I.

    2001-03-01

    Full Text Available To recycle and reutilise waste materials and find definite applications for their use, it is necessary to have a deep knowledge of them. The aim of this study is to study the possibility of using waste materials from calcined clay, actually ceramic tile, once crushed and grounded, as pozzolanic material. For this purpose, different tests are carried out in order to establish the pozzolanic activity of this material. At the same time, these results are compared to those of other industrial by-products, fly ash and silica fume, which are pozzolanic materials usually employed to elaborate mortars and concretes.

    Para llevar a cabo labores encaminadas al reciclado y revalorización de residuos es necesario un conocimiento profundo de los mismos, de forma que se busquen aplicaciones concretas de uso. El objetivo de este estudio es investigar la posibilidad de utilizar materiales de desecho procedentes de arcilla cocida, concretamente teja cerámica, una vez triturada y molida, como puzolana. Para ello, se efectúan diferentes ensayos dirigidos a establecer la actividad puzolanica del material. A su vez, estos resultados son comparados con otros residuos industriales, ceniza volante y humo de sílice, habituales en la elaboración de morteros y hormigones.

  20. Draft environmental assessment -- Closure of the Waste Calcining Facility (CPP-633), Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Braun, J.B.; Irving, J.S.; Staley, C.S.; Stanley, N.

    1996-04-01

    The DOE-Idaho Operations Office has prepared an environmental assessment (EA) to analyze the environmental impacts of closing the Waste Calcining Facility (WCF) at the Idaho National Engineering Laboratory (INEL). The purpose of the action is to reduce the risk of radioactive exposure and release of radioactive and hazardous constituents and eliminate the need for extensive long-term surveillance and maintenance. DOE has determined that the closure is needed to reduce these risks to human health and the environment and to comply with Resource Conservation and Recovery Act requirements. The WCF closure project is described in the DOE Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement (Programmatic EIS). DOE determined in the Programmatic EIS Record of Decision (ROD) that certain actions would be implemented and other actions deferred. The EA examined the potential environmental impacts of the proposed action and evaluated reasonable alternatives, including the no action alternative in accordance with the Council on Environmental Quality Regulations. Based on the analysis in the EA, the action will not have a significant effect on the human environment.

  1. Regeneration of Waste Edible Oil by the Use of Virgin and Calcined Magnesium Hydroxide as Adsorbents.

    Science.gov (United States)

    Ogata, Fumihiko; Kawasaki, Naohito

    2016-01-01

    In this study, we prepared virgin (S, L) and calcined (S-380, S-1000, L-380, L-1000) magnesium hydroxide for regeneration of waste edible oil. Deterioration of soybean oil, rapeseed oil, and olive oil was achieved by heat and aeration treatment. The properties of the different adsorbents were investigated using specific surface area measurements, scanning electron microscopy, X-ray diffraction analysis, thermogravimetric-differential thermal analysis, and surface pH measurement. Moreover, the relationship between the changes in acid value (AV) and carbonyl value (CV) and the adsorbent properties were evaluated. The specific surface areas of S-380 and L-380 were greater than that of other adsorbents. In addition, the XRD results show that S-380 and L-380 contain both magnesium hydroxide and magnesium oxide structures. The decreases in AV and CV using S-380 and L-380 were greater than achieved using other adsorbents. The correlation coefficients between the decrease in AV and CV and specific surface area were 0.947 for soybean oil, 0.649 for rapeseed oil, and 0.773 for olive oil, respectively. The results obtained in this study suggest that a physical property of the adsorbent, namely specific surface area, was primarily responsible for the observed decreases in AV and CV. Overall, the results suggest that S-380 and L-380 are useful for the regeneration of waste edible oil.

  2. 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.

  3. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

  4. Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, W.M. [Center for Nuclear Waste Regulations Analyses, San Antonio, TX (United States); Kovach, L.A. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-09-01

    A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research related to geologic disposal of HLW.

  5. Evaluation of high-level waste pretreatment processes with an approximate reasoning model

    Energy Technology Data Exchange (ETDEWEB)

    Bott, T.F.; Eisenhawer, S.W.; Agnew, S.F.

    1999-04-01

    The development of an approximate-reasoning (AR)-based model to analyze pretreatment options for high-level waste is presented. AR methods are used to emulate the processes used by experts in arriving at a judgment. In this paper, the authors first consider two specific issues in applying AR to the analysis of pretreatment options. They examine how to combine quantitative and qualitative evidence to infer the acceptability of a process result using the example of cesium content in low-level waste. They then demonstrate the use of simple physical models to structure expert elicitation and to produce inferences consistent with a problem involving waste particle size effects.

  6. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  7. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-06-02

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  8. Structural integrity and potential failure modes of hanford high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Han, F.C.

    1996-09-30

    Structural Integrity of the Hanford High-Level Waste Tanks were evaluated based on the existing Design and Analysis Documents. All tank structures were found adequate for the normal operating and seismic loads. Potential failure modes of the tanks were assessed by engineering interpretation and extrapolation of the existing engineering documents.

  9. Structural Integrity Program for INTEC Calcined Solids Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Bryant

    2008-08-30

    This report documents the activities of the structural integrity program at the Idaho Nuclear Technology and Engineering Center relevant to the high-level waste Calcined Solids Storage Facilities and associated equipment, as required by DOE M 435.1-1, 'Radioactive Waste Management Manual'. Based on the evaluation documented in this report, the Calcined Solids Storage Facilities are not leaking and are structurally sound for continued service. Recommendations are provided for continued monitoring of the Calcined Solids Storage Facilities.

  10. 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.

  11. Overview of Hanford Site High-Level Waste Tank Gas and Vapor Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Huckaby, James L.; Mahoney, Lenna A.; Droppo, James G.; Meacham, Joseph E.

    2004-08-31

    Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-level waste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

  12. A Study on Site Selecting for National Project including High Level Radioactive Waste Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kilyoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Many national projects are stopped since sites for the projects are not determined. The sites selections are hold by NIMBY for unpleasant facilities or by PYMFY for preferable facilities among local governments. The followings are the typical ones; NIMBY projects: high level radioactive waste disposal, THAAD, Nuclear power plant(NPP), etc. PIMFY projects: South-east new airport, KTX station, Research center for NPP decommission, etc. The site selection for high level radioactive waste disposal is more difficult problem, and thus government did not decide and postpone to a dead end street. Since it seems that there is no solution for site selection for high level radioactive waste disposal due to NIMBY among local governments, a solution method is proposed in this paper. To decide a high level radioactive waste disposal, the first step is to invite a bid by suggesting a package deal including PIMFY projects such as Research Center for NPP decommission. Maybe potential host local governments are asked to submit sealed bids indicating the minimum compensation sum that they would accept the high level radioactive waste disposal site. If there are more than one local government put in a bid, then decide an adequate site by considering both the accumulated PESS point and technical evaluation results. By considering how fairly preferable national projects and unpleasant national projects are distributed among local government, sites selection for NIMBY or PIMFY facilities is suggested. For NIMBY national projects, risk, cost benefit analysis is useful and required since it generates cost value to be used in the PESS. For many cases, the suggested method may be not adequate. However, similar one should be prepared, and be basis to decide sites for NIMBY or PIMFY national projects.

  13. 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.

  14. Natural geochemical analogues of the near field of high-level nuclear waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Apps, J.A. [Lawrence Berkeley Lab., CA (United States)

    1995-09-01

    United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100{degrees}C and could reach 250{degrees}C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields.

  15. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  16. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

  17. Reference design and operations for deep borehole disposal of high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

    2011-10-01

    A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    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.

  20. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

  1. Bin Set 1 Calcine Retrieval Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    R. D. Adams; S. M. Berry; K. J. Galloway; T. A. Langenwalter; D. A. Lopez; C. M. Noakes; H. K. Peterson; M. I. Pope; R. J. Turk

    1999-10-01

    At the Department of Energy's Idaho Nuclear Technology and Engineering Center, as an interim waste management measure, both mixed high-level liquid waste and sodium bearing waste have been solidified by a calculation process and are stored in the Calcine Solids Storage Facilities. This calcined product will eventually be treated to allow final disposal in a national geologic repository. The Calcine Solids Storage Facilities comprise seven ''bit sets.'' Bin Set 1, the first to be constructed, was completed in 1959, and has been in service since 1963. It is the only bin set that does not meet current safe-shutdown earthquake seismic criteria. In addition, it is the only bin set that lacks built-in features to aid in calcine retrieval. One option to alleviate the seismic compliance issue is to transport the calcine from Bin Set 1 to another bin set which has the required capacity and which is seismically qualified. This report studies the feasibility of retrieving the calcine from Bi n Set 1 and transporting it into Bin Set 6 which is located approximately 650 feet away. Because Bin Set 1 was not designed for calcine retrieval, and because of the high radiation levels and potential contamination spread from the calcined material, this is a challenging engineering task. This report presents preconceptual design studies for remotely-operated, low-density, pneumatic vacuum retrieval and transport systems and equipment that are based on past work performed by the Raytheon Engineers and Constructors architectural engineering firm. The designs presented are considered feasible; however, future development work will be needed in several areas during the subsequent conceptual design phase.

  2. Development of anodic stripping voltametry for the determination of palladium in high level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Bhardwaj, T. K. [North Carolina State University, Raleigh (United States); Sharma, H. S.; Affarwal, S. K. [Bhabha Atomic Research Centre, Mumbai (India); Jain, P. C. [Meerut College, Meerut (India)

    2012-12-15

    Deposition potential, deposition time, square wave frequency, rotation speed of the rotating disc electrode, and palladium concentration were studied on a Glassy Carbon Electrode (GCE) in 0.01M HCl for the determination of palladium in High Level Nuclear Waste (HLNW) by anodic stripping voltammetry. Experimental conditions were optimized for the determination of palladium at two different, 10-8 and 10-7 M, levels. Error and standard deviation of this method were under 1% for all palladium standard solutions. The developed technique was successfully applied as a subsidiary method for the determination of palladium in simulated high level nuclear waste with very good precision and high accuracy (under 1 % error and standard deviation).

  3. Development of site suitability criteria for the high level waste repository for Lawrence Livermore Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    Results of our mining, geological and geotechnical studies provided in support of the development of site suitability criteria for the high level waste repository are presented. The primary purpose of the work was the identification and development of appropriate geotechnical descriptors and coefficients required for the Site Suitability Repository Model. This model was developed by The Analytic Sciences Corporation (TASC) of Reading, Massachusetts and is not described in this report.

  4. Removal of Aerosol Particles Generated from Vitrification Process for High-Level Liquid Wastes

    OpenAIRE

    加藤 功

    1990-01-01

    The vitrification technology has been developed for the high-level liquid waste (HLLW) from reprocessing nuclear spent fuel in PNC. The removal performance of the aerosol particles generated from the melting process was studied in a nonradioactive full-scale mock-up test facility (MTF). The off-gas treatment system consists of submerged bed scrubber (SBS), venturi scrubber, NOx absorber, high efficiency mist eliminater (HEME). Deoomtamination factors (DFs) were derived from the mass ratio of ...

  5. Geological aspects of the high level waste and spent fuel disposal programme in Slovakia

    Energy Technology Data Exchange (ETDEWEB)

    Matej, Gedeon; Milos, Kovacik; Jozef, Hok [Geological Survey of Slovak Republic, Bratislava (Slovakia)

    2001-07-01

    An autonomous programme for development of a deep geological high level waste and spent fuel disposal began in 1996. One of the most important parts in the programme is siting of the future deep seated disposal. Geological conditions in Slovakia are complex due to the Alpine type tectonics that formed the geological environment during Tertiary. Prospective areas include both crystalline complexes (tonalites, granites, granodiorites) and Neogene (Miocene) argillaceous complexes. (author)

  6. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Odriscoll, R; Allan Barnes, A; Jim Coleman, J; Timothy Glover, T; Robert Hopkins, R; Dan Iverson, D; Jeff Leita, J

    2008-01-15

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing high level waste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the high level waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  8. Characteristics of high-level radioactive waste forms for their disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

    2000-12-01

    In order to develop a deep geological repository for a high-level radioactive waste coming from reprocessing of spent nuclear fuels discharged from our domestic nuclear power plants, the the required characteristics of waste form are dependent upon a solidifying medium and the amount of waste loading in the medium. And so, by the comparative analysis of the characteristics of various waste forms developed up to the present, a suitable medium is recommended.The overall characteristics of the latter is much better than those of the former, but the change of the properties due to an amorphysation by radiation exposure and its thermal expansion has not been clearly identified yet. And its process has not been commercialized. However, the overall properties of the borosilicate glass waste forms are acceptable for their disposal, their production cost is reasonable and their processes have already been commercialized. And plenty informations of their characteristics and operational experiences have been accumulated. Consequently, it is recommended that a suitable medium solidifying the HLW is a borosilicate glass and its composition for the identification of a reference waste form would be based on the glass frit of R7T7.

  9. Evaluation of long-term irradiation field in geological disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, H.; Nishimura, K.; Neyama, A. [Computer Software Development Co., Ltd., Shinjyuku, Tokyo (Japan); Naito, M.; Ohi, T.; Ishihara, Y. [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    Vitrified high-level radioactive waste (HLW) is subject to alpha, beta, gamma and neutron irradiation as a result of radionuclide-decay. Radiation can cause chemical and physical effects on HLW geological disposal system, in particular, engineered barrier system (EBS) which consists of vitrified waste, overpack container and surrounded buffer material. Alpha and beta radiation can be shielded completely by the overpack as long as it retains its containment function. Gamma and neutron radiation, on the other hand, will penetrate the overpack, and then enter the buffer material and the host rock. To assess radiation effects within the EBS for long time, it is essential to evaluate the evolution of irradiation field, quantitatively. Thus, radiation transport calculations were done to obtain dose rate, irradiation dose and absorbed dose in the irradiation field. In these calculations, vitrified waste, overpack, buffer and host rock were modeled with a same concentric cylinder. (author)

  10. Can Sisyphus succeed? Getting U.S. high-level nuclear waste into a geological repository.

    Science.gov (United States)

    North, D Warner

    2013-01-01

    The U.S. government has the obligation of managing the high-level radioactive waste from its defense activities and also, under existing law, from civilian nuclear power generation. This obligation is not being met. The January 2012 Final Report from the Blue Ribbon Commission on America's Nuclear Future provides commendable guidance but little that is new. The author, who served on the federal Nuclear Waste Technical Review Board from 1989 to 1994 and subsequently on the Board on Radioactive Waste Management of the National Research Council from 1994 to 1999, provides a perspective both on the Commission's recommendations and a potential path toward progress in meeting the federal obligation. By analogy to Sisyphus of Greek mythology, our nation needs to find a way to roll the rock to the top of the hill and have it stay there, rather than continuing to roll back down again. © 2012 Society for Risk Analysis.

  11. Spring 2009 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post-Closure Permit for the INTEC Waste Calcining Facility at the INL Site

    Energy Technology Data Exchange (ETDEWEB)

    Boehmer, Ann M.

    2009-05-31

    The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under and approved Hazardous Waste Management Act/Resource Conservation and Recovery Act Closure plan. Vessels and spaces were grouted and then covered with a concrete cap. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment.

  12. Radiant-heat spray-calcination process for the solid fixation of radioactive waste. Part 1, Non-radioactive pilot unit

    Energy Technology Data Exchange (ETDEWEB)

    Allemann, R.T.; Johnson, B.M. Jr.

    1960-11-14

    The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

  13. Next Generation Extractants for Cesium Separation from High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, Bruce A [ORNL; Bazelaire, Eve [ORNL; Bonnesen, Peter V [ORNL; Custelcean, Radu [ORNL; Delmau, Laetitia Helene [ORNL; Ditto, Mary E [ORNL; Engle, Nancy L [ORNL; Gorbunova, Maryna [ORNL; Haverlock, Tamara [ORNL; Levitskaia, Tatiana G. [Pacific Northwest National Laboratory (PNNL); Bartsch, Richard A. [Texas Tech University, Lubbock; Surowiec, Malgorzata A. [Texas Tech University, Lubbock; Marquez, Manuel [University of Texas; Zhou, Hui [Texas Tech University, Lubbock

    2006-01-01

    This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed bear upon cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit applied projects funded by the USDOE Office of Environmental Management to clean up sites such as the Savannah River Site (SRS), Hanford, and the Idaho National Environmental and Engineering Laboratory. The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high-level tank waste. Disposal of high-level waste is horrendously expensive, in large part because the actual radioactive matter in underground waste tanks at various USDOE sites has been diluted over 1000-fold by ordinary inorganic chemicals. To vitrify the entire mass of the high-level waste would be prohibitively expensive. Accordingly, an urgent need has arisen for technologies to remove radionuclides such as {sup 137}Cs from the high-level waste so that the bulk of it may be diverted to cheaper low-level waste forms and cheaper storage. To address this need in part, chemical research at Oak Ridge National Laboratory (ORNL) has focused on calixcrown extractants, molecules that combine a crown ether with a calixarene. This hybrid possesses a cavity that is highly complementary for the Cs{sup +} ion vs. the Na+ ion, making it possible to cleanly separate cesium from wastes that contain 10,000- to 1,000,000-fold higher concentrations of sodium. Previous EMSP results in Project 55087 elucidated the underlying extraction

  14. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

  15. Creating a proper safety culture at the Hanford Site low- and high-level waste vitrification plant projects

    Energy Technology Data Exchange (ETDEWEB)

    Baide, D.G.; Herborn, D.I.

    1994-05-01

    The United States has been engaged in defense nuclear activities at the Hanford Site for the past 50 years. To date, no high-level waste and only 3,800 m{sup 3} of low-level waste have been processed for final disposal. By the anticipated start of low-level waste processing operations in the year 2005, approximately 215,000 m{sup 3} of low-level waste will be in underground storage tanks (90% of the total tank waste in storage). Similarly, approximately 25,000 m{sup 3} of high-level waste will be in underground storage by the anticipated start of high-level waste processing operations in the year 2009 (10% of the total tank waste in storage).

  16. Risk perception on management of nuclear high-level and transuranic waste storage

    Energy Technology Data Exchange (ETDEWEB)

    Dees, Lawrence A. [Colorado Christian Univ., Lakewood, CO (United States)

    1994-08-15

    The Department of Energy`s program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

  17. Control of high level radioactive waste-glass melters. Part 5, Modelling of complex redox effects

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.; Choi, A.S.

    1991-12-31

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs.

  18. Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    B. A. Staples; T. P. O' Holleran

    1999-05-01

    The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

  19. Methods of calculating the post-closure performance of high-level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Ross, B. (ed.)

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

  20. ITP. FOR: A code to calculate thermal transients in High Level Waste Tanks

    Energy Technology Data Exchange (ETDEWEB)

    Kielpinski, A.L.

    1992-10-01

    A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code's applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code's capabilities have generated some interest to date, the present report is presented as interim documentation of the code's mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code.

  1. ITP.FOR: A code to calculate thermal transients in High Level Waste Tanks

    Energy Technology Data Exchange (ETDEWEB)

    Kielpinski, A.L.

    1992-10-01

    A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code`s applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code`s capabilities have generated some interest to date, the present report is presented as interim documentation of the code`s mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code.

  2. Policy Requirements and Factors of High-Level Radioactive Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Myoung; Jeong, J. Y.; Ha, K. M. [Korea Energy Technology and Emergency Management Institute, Seoul (Korea, Republic of)

    2007-06-15

    Recently, the need of high-level radioactive waste policy including spent fuel management becomes serious due to the rapid increase in oil price, the nationalism of natural resources, and the environmental issues such as Tokyo protocol. Also, the policy should be established urgently to prepare the saturation of on-site storage capacity of spent fuel, the revision of 'Agreement for Cooperation-Concerning Civil Uses of Atomic Energy' between Korea and US, the anxiety for nuclear weapon proliferation, and R and D to reduce the amount of waste to be disposed. In this study, we performed case study of US, Japan, Canada and Finland, which have special laws and plans/roadmaps for high-level waste management, to draw the policy requirements to be considered in HLW management. Also, we reviewed social conflict issues experienced in our society, and summarized the factors affecting the political and social environment. These policy requirements and factors summarized in this study should be considered seriously in the process for public consensus and the policy making regarding HLW management. Finally, the following 4 action items were drawn to manage HLW successfully : - Continuous and systematic R and D activities to obtain reliable management technology - Promoting companies having specialty in HLW management - Nurturing experts and workforce - Drive the public consensus process

  3. High Level Waste System Impacts from Small Column Ion Exchange Implementation

    Energy Technology Data Exchange (ETDEWEB)

    McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

    2005-08-18

    The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were

  4. Potential Application Of Radionuclide Scaling Factors To High Level Waste Characterization

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-30

    Production sources, radiological properties, relative solubilities in waste, and laboratory analysis techniques for the forty-five radionuclides identified in Hanford's Waste Treatment and Immobilization Plant (WTP) Feed Acceptance Data Quality Objectives (DQO) document are addressed in this report. Based on Savannah River Site (SRS) experience and waste characteristics, thirteen of the radionuclides are judged to be candidates for potential scaling in High Level Waste (HLW) based on the concentrations of other radionuclides as determined through laboratory measurements. The thirteen radionuclides conducive to potential scaling are: Ni-59, Zr-93, Nb-93m, Cd-113m, Sn-121m, Sn-126, Cs-135, Sm-151, Ra-226, Ra-228, Ac-227, Pa-231, and Th-229. The ability to scale radionuclides is useful from two primary perspectives: 1) it provides a means of checking the radionuclide concentrations that have been determined by laboratory analysis; and 2) it provides a means of estimating radionuclide concentrations in the absence of a laboratory analysis technique or when a complex laboratory analysis technique fails. Along with the rationale for identifying and applying the potential scaling factors, this report also provides examples of using the scaling factors to estimate concentrations of radionuclides in current SRS waste and into the future. Also included in the report are examples of independent laboratory analysis techniques that can be used to check results of key radionuclide analyses. Effective utilization of radionuclide scaling factors requires understanding of the applicable production sources and the chemistry of the waste. As such, the potential scaling approaches identified in this report should be assessed from the perspective of the Hanford waste before reaching a decision regarding WTP applicability.

  5. Application of calcined waste fish (Labeo rohita) scale as low-cost heterogeneous catalyst for biodiesel synthesis.

    Science.gov (United States)

    Chakraborty, R; Bepari, S; Banerjee, A

    2011-02-01

    This paper explores the feasibility of converting waste Rohu fish (Labeo rohita) scale into a high-performance, reusable, low-cost heterogeneous catalyst for synthesis of biodiesel from soybean oil. The thermo-gravimetric analysis (TGA) and X-ray diffraction (XRD) analysis revealed that a significant portion of the main component of fish scale i.e. HAP (hydroxyapatite) could be transformed into β-tri-calcium phosphate when calcined above 900°C for 2 h. Scanning Electron Microscopy (SEM) morphology studies of the calcined scale depicted a fibrous layer of porous structure; while a BET surface area of 39 m(2)/g was measured. Response surface methodology (RSM) was employed to determine the optimal parametric conditions viz. methanol/oil molar ratio, 6.27:1, calcination temperature, 997.42°C and catalyst concentration, 1.01 wt.% of oil corresponding to a maximum FAME yield of 97.73%. Reusability results confirmed that the prepared catalyst could be reemployed up to six times, procreating a potentially applicable avenue in biodiesel synthesis. Copyright © 2010 Elsevier Ltd. All rights reserved.

  6. Calcined Eggshell Waste for Mitigating Soil Antibiotic-Resistant Bacteria/Antibiotic Resistance Gene Dissemination and Accumulation in Bell Pepper.

    Science.gov (United States)

    Ye, Mao; Sun, Mingming; Feng, Yanfang; Li, Xu; Schwab, Arthur P; Wan, Jinzhong; Liu, Manqiang; Tian, Da; Liu, Kuan; Wu, Jun; Jiang, Xin

    2016-07-13

    The combined accumulation of antibiotics, heavy metals, antibiotic-resistant bacteria (ARB)/antibiotic resistance genes (ARGs) in vegetables has become a new threat to human health. This is the first study to investigate the feasibility of calcined eggshells modified by aluminum sulfate as novel agricultural wastes to impede mixed contaminants from transferring to bell pepper (Capsicum annuum L.). In this work, calcined eggshell amendment mitigated mixed pollutant accumulation in bell pepper significantly, enhanced the dissipation of soil tetracycline, sulfadiazine, roxithromycin, and chloramphenicol, decreased the water-soluble fractions of antibiotics, and declined the diversity of ARB/ARGs inside the vegetable. Moreover, quantitative polymerase chain reaction analysis detected that ARG levels in the bell pepper fruits significantly decreased to 10(-10) copies/16S copies, indicating limited risk of ARGs transferring along the food chain. Furthermore, the restoration of soil microbial biological function suggests that calcined eggshell is an environmentally friendly amendment to control the dissemination of soil ARB/ARGs in the soil-vegetable system.

  7. Southern routes for high-level radioactive waste: Agencies, contacts, and designations

    Energy Technology Data Exchange (ETDEWEB)

    1991-05-01

    The Southern Routes for High-Level Radioactive Waste: Agencies, Contacts and Designations is a compendium of sixteen southern states' routing programs for the transportation of high-level radioactive materials. The report identifies the state-designated routing agencies as defined under 49 Code of Federal Regulations (CFR) Part 171 and provides a reference to the source and scope of the agencies' rulemaking authority. Additionally, the state agency and contact designated by the state's governor to receive advance notification and shipment routing information under 10 CFR Parts 71 and 73 are also listed. This report also examines alternative route designations made by southern states and the lessons that were learned from the designation process.

  8. Southern routes for high-level radioactive waste: Agencies, contacts, and designations

    Energy Technology Data Exchange (ETDEWEB)

    1991-05-01

    The Southern Routes for High-Level Radioactive Waste: Agencies, Contacts and Designations is a compendium of sixteen southern states` routing programs for the transportation of high-level radioactive materials. The report identifies the state-designated routing agencies as defined under 49 Code of Federal Regulations (CFR) Part 171 and provides a reference to the source and scope of the agencies` rulemaking authority. Additionally, the state agency and contact designated by the state`s governor to receive advance notification and shipment routing information under 10 CFR Parts 71 and 73 are also listed. This report also examines alternative route designations made by southern states and the lessons that were learned from the designation process.

  9. Validation of the Performance of High-level Waste Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Park, J. H.; Lee, J. O. (and others)

    2007-06-15

    The experimental researches to validate the integrity and safety of high-level waste disposal system were carried out. The studies on the construction of KURT, and the site rock characteristics were conducted. Thermal-hydro-mechanical behavior of engineered barrier system was investigated using the engineering-scale test facility. The migration and retardation of radionuclide through the rock fracture under anaerobic and reducing condition were studied. The distribution coefficients of radionuclides onto granite, the rock matrix diffusion coefficients, and the gap and grain boundary inventories of spent fuel were measured.

  10. Analysis of functional criteria for buffer material in a high-level waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Chun, Kwan Sik; Lee, Jae Owan; Kang, Mun Ja

    1997-11-01

    This study is intended to analyze the requirements of buffer material that is one of the major components of the engineered barriers in a high-level waste repository. Based on the results, it is intended to suggest the quantitative functional criteria that is necessary to establish the preliminary concept for the domestic geological repository. The criteria are composed of seven major items, such as hydraulic conductivity, retardation capacity, swelling potential and swelling pressure, thermal conductivity, longevity, organic carbon content, and mechanical properties. (author). 87 refs., 12 tabs., 3 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12

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

  12. High level nuclear waste repository in salt: Sealing systems status and planning report: Draft report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-09-01

    This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs.

  13. Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

  14. Development of a thermal transient calculational tool for High Level Waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Kielpinski, A.L.

    1994-06-01

    Thermal design constraints exist on the processing operations in the High Level Waste (HLW) tanks of the Savannah River Site (SRS). A FORTRAN computer code was developed to provide a simple, fast, and reasonably accurate analysis tool for plant operation design. The code computes a lumped transient temperature for the liquid contents of a waste tank by modeling the liquid (slurry), the vapor space above it, the tank wall, and the cooling air outside of the tank. Results for a typical processing cycle of several months` duration can be obtained in 2--4 minutes CPU time on a VAX computer. This paper discusses the code`s mathematical models, presents model results for a typical HLW process schedule, and compares the code predictions with operations data.

  15. Conceptual modular description of the high-level waste management system for system studies model development

    Energy Technology Data Exchange (ETDEWEB)

    McKee, R.W.; Young, J.R.; Konzek, G.J.

    1992-08-01

    This document presents modular descriptions of possible alternative components of the federal high-level radioactive waste management system and the procedures for combining these modules to obtain descriptions for alternative configurations of that system. The 20 separate system component modules presented here can be combined to obtain a description of any of the 17 alternative system configurations (i.e., scenarios) that were evaluated in the MRS Systems Studies program (DOE 1989a). First-approximation descriptions of other yet-undefined system configurations could also be developed for system study purposes from this database. The descriptions include, in a modular format, both functional descriptions of the processes in the waste management system, plus physical descriptions of the equipment and facilities necessary for performance of those functions.

  16. A risk governance approach for high-level waste in Belgium: A process appraisal

    Energy Technology Data Exchange (ETDEWEB)

    Laes, Erik (Flemish Inst. for Technological Research (VITO)/Univ. of Antwerp, Div. Transition Energy and Environment, Mol (Belgium)), e-mail: erik.laes@vito.be; Eggermont, Gilbert (Free Univ. of Brussels (VUB), Brussels (Belgium)); Bombaerts, Gunter ((Belgium))

    2010-09-15

    The Belgian nuclear waste management organisation (NIRAS-ONDRAF) has recently started up a public debate on the strategic waste management options for the intermediate- and high-level radioactive waste (cat. B and C waste). This public debate takes place in the context of a (mandatory) strategic environmental impact assessment (SEA) procedure. The paper proposes a critical investigation of four interrelated aspects of this procedure from the point of view of 'good governance': assessment of the remaining uncertainties, guardianship of the democratic process, the organisation of expertise and the interpretation of transgenerational ethics and distributive justice in the new crisis context of globalization and failure of electricity liberalisation. We argue that - in spite of the overall soundness of the geological disposal option - many uncertainties remain: a new technical concept needs to be demonstrated and international financial management needs to be organised. On the process side we argue that although NIRAS-ONDRAF can take up a role as initiator of a public participation process, it can hardly act as a guardian of this process. The debate must be lifted above the local level, opened up to new actors with an active role of the safety authorities and guarded by a non-involved organisation. A condition for success is the creation of critical awareness and the capacity to manage controversy in future with critical expertise. Referring to the RISCOM model for transparent risk communication, we suggest some improvements to the process that is currently taking place

  17. Department of Energy pretreatment of high-level and low-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    McGinnis, C.P.; Hunt, R.D.

    1995-12-31

    The remediation of the 1 {times} 10{sup 8} gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE`s greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste.

  18. Nuclear waste solidification

    Science.gov (United States)

    Bjorklund, William J.

    1977-01-01

    High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

  19. Preconceptual design study for solidifying high-level waste: Appendices A, B and C West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Hill, O.F. (comp.)

    1981-04-01

    This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass.

  20. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  1. Research and development plans for disposal of high-level and transuranic wastes

    Energy Technology Data Exchange (ETDEWEB)

    Bartlett, J.W.; Platt, A.M.

    1978-09-01

    This plan recommends a 20-year, 206 million (1975 $'s) R and D program on geologic structures in the contiguous U.S. and on the midplate Pacific seabed with the objective of developing an acceptable method for disposal of commercial high-level and transuranic wastes by 1997. No differentiation between high-level and transuranic waste disposal is made in the first 5 years of the program. A unique application of probability theory to R and D planning establishes, at a 95% confidence level, that the program objective will be met if at least fifteen generic options and five specific disposal sites are explored in detail and at least two pilot plants are constructed and operated. A parallel effort on analysis and evaluation maximizes information available for decisions on the acceptability of the disposal techniques. Based on considerations of technical feasibility, timing and technical risk, the other disposal concepts, e.g., ice sheets, partitioning, transmutation and space disposal cited in BNWL-1900 are not recommended for near future R and D.

  2. Thirty Years of Social Science Research on High-Level Nuclear Waste: Achievements and Future Challenges

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, Barry D. (Dept. of Social Sciences, Michigan Technological Univ., Houghton (United States)), e-mail: bdsolomo@mtu.edu; Andren, Mats; Strandberg, Urban (Center for Public Sector Research, Univ. of Goeteborg, Goeteborg (Sweden))

    2010-09-15

    Research on high-level nuclear waste management has focused on technical and scientific issues since the U.S. National Academy of Sciences first studied the problem in the mid 1950s and recommended long-term disposal in deep salt formations. In this review, we trace the development of the problem's definition and its associated research since socioeconomic, political and policy issues were first given consideration and nuclear waste management became recognized as more than a technical issue. Three time periods are identified. First, from the mid 1970s to early 1980s, initial research explored institutional dimensions of nuclear waste, including ethics. The second period began in the early 1980s with a concerted effort to solve the problem and site nuclear waste repositories, and ended in the mid 1990s with minimal progress in the U.S. and general stalemate in Asia and Europe (with the notable exception of Sweden). This phase accelerated research on risk perception and stigma of nuclear waste, and elevated a focus on public trust. Great attention was given to repository siting conflicts, while minimal attention was placed on ethics, equity, political systems, and public participation. The last period, since the mid 1990s, has been characterized by continuing political stalemate and increased attention to public participation, political systems and international solutions. Questions of ethics have been given renewed attention, while research on risk perceptions and siting conflicts continues. We frame these periods in a broader context of the shifting role of applied social scientists. The paper concludes with a general discussion of this research area and prospects for future research

  3. Granite disposal of U.S. high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A.; Mariner, Paul E.; Lee, Joon H.; Hardin, Ernest L.; Goldstein, Barry; Hansen, Francis D.; Price, Ronald H.; Lord, Anna Snider

    2011-08-01

    This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site

  4. High level liquid waste solidification using a ''Cold' crucible induction melter

    Energy Technology Data Exchange (ETDEWEB)

    Demine, A.V.; Krylova, N.V.; Polyektov, P.P.; Shestoperov, I.N.; Smelova, T.V. [SSC RF VNIINM, Moscow (Russian Federation); Gorn, V.F.; Medvedev, G.M. [IA ' ' MAYAK' ' , Ozersk (Russian Federation)

    2000-07-01

    At the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from the spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a 'cold' crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter. This paper describes the features of a process and equipment for two-stage HLLW solidification technology using a 'cold' crucible induction melter (CCIM) and its advantages compared to a one stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporated and induction cold-crucible melter currently operational at the IA 'Mayak' facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak - first to produce borosilicate glass waste forms then mineral-like waste forms. (authors)

  5. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-04-01

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

  6. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Wurm, K.J.; Miller, N.E.

    1982-11-01

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

  7. Site Selection and Geological Research Connected with High Level Waste Disposal Programme in the Czech Republic

    Energy Technology Data Exchange (ETDEWEB)

    Tomas, J.

    2003-02-25

    Attempts to solve the problem of high-level waste disposal including the spent fuel from nuclear power plants have been made in the Czech Republic for over the 10 years. Already in 1991 the Ministry of Environment entitled The Czech Geological Survey to deal with the siting of the locality for HLW disposal and the project No. 3308 ''The geological research of the safe disposal of high level waste'' had started. Within this project a sub-project ''A selection of perspective HLW disposal sites in the Bohemian Massif'' has been elaborated and 27 prospective areas were identified in the Czech Republic. This selection has been later narrowed to 8 areas which are recently studied in more detail. As a parallel research activity with siting a granitic body Melechov Massif in Central Moldanubian Pluton has been chosen as a test site and the 1st stage of research i.e. evaluation and study of its geological, hydrogeological, geophysical, tectonic and structural properties has been already completed. The Melechov Massif was selected as a test site after the recommendation of WATRP (Waste Management Assessment and Technical Review Programme) mission of IAEA (1993) because it represents an area analogous with the host geological environment for the future HLW and spent fuel disposal in the Czech Republic, i.e. variscan granitoids. It is necessary to say that this site would not be in a locality where the deep repository will be built, although it is a site suitable for oriented research for the sampling and collection of descriptive data using up to date and advanced scientific methods. The Czech Republic HLW and spent fuel disposal programme is now based on The Concept of Radioactive Waste and Spent Nuclear Fuel Management (''Concept'' hereinafter) which has been prepared in compliance with energy policy approved by Government Decree No. 50 of 12th January 2000 and approved by the Government in May 2002. Preparation of

  8. Shale disposal of U.S. high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Sassani, David Carl; Stone, Charles Michael; Hansen, Francis D.; Hardin, Ernest L.; Dewers, Thomas A.; Martinez, Mario J.; Rechard, Robert Paul; Sobolik, Steven Ronald; Freeze, Geoffrey A.; Cygan, Randall Timothy; Gaither, Katherine N.; Holland, John Francis; Brady, Patrick Vane

    2010-05-01

    This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within

  9. Development of pyrometallurgical partitioning technology for TRU in high level radioactive wastes. Vitrification process for salt wastes

    Energy Technology Data Exchange (ETDEWEB)

    Sakamura, Yoshiharu; Inoue, Tadashi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Shimizu, Takafumi; Kobayashi, Kuniaki

    1997-12-31

    A vitrification process for chloride wastes generated in the pyrometallurgical partitioning of TRUs from high level radioactive wastes is being developed. In the process, chlorides are reduced to metals by molten salt electrolysis. The metals are oxidized by air and then vitrified. Lithium metal and chlorine gas are recycled. The behaviors of lithium, sodium and fission products during molten salt electrolysis were studied by using various compositions of salts and cathode materials. It was shown that every metal can be recovered into a liquid lead cathode, and that a liquid cadmium cathode and a solid cathode are suitable for recovering lithium and sodium metal, respectively. Based on the experimental results the process flow sheet was discussed. (author)

  10. DOUBLE SHELL TANK (DST) INTEGRITY PROJECT HIGH LEVEL WASTE CHEMISTRY OPTIMIZATION

    Energy Technology Data Exchange (ETDEWEB)

    WASHENFELDER DJ

    2008-01-22

    The U.S. Department of Energy's Office (DOE) of River Protection (ORP) has a continuing program for chemical optimization to better characterize corrosion behavior of High-Level Waste (HLW). The DOE controls the chemistry in its HLW to minimize the propensity of localized corrosion, such as pitting, and stress corrosion cracking (SCC) in nitrate-containing solutions. By improving the control of localized corrosion and SCC, the ORP can increase the life of the Double-Shell Tank (DST) carbon steel structural components and reduce overall mission costs. The carbon steel tanks at the Hanford Site are critical to the mission of safely managing stored HLW until it can be treated for disposal. The DOE has historically used additions of sodium hydroxide to retard corrosion processes in HLW tanks. This also increases the amount of waste to be treated. The reactions with carbon dioxide from the air and solid chemical species in the tank continually deplete the hydroxide ion concentration, which then requires continued additions. The DOE can reduce overall costs for caustic addition and treatment of waste, and more effectively utilize waste storage capacity by minimizing these chemical additions. Hydroxide addition is a means to control localized and stress corrosion cracking in carbon steel by providing a passive environment. The exact mechanism that causes nitrate to drive the corrosion process is not yet clear. The SCC is less of a concern in the newer stress relieved double shell tanks due to reduced residual stress. The optimization of waste chemistry will further reduce the propensity for SCC. The corrosion testing performed to optimize waste chemistry included cyclic potentiodynamic volarization studies. slow strain rate tests. and stress intensity factor/crack growth rate determinations. Laboratory experimental evidence suggests that nitrite is a highly effective:inhibitor for pitting and SCC in alkaline nitrate environments. Revision of the corrosion control

  11. RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, F.; Edwards, T.

    2010-06-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and

  12. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Bedrossian, P.J.; McCright, R.D.

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological respository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environmental (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

  13. Phase chemistry and radionuclide retention of high level radioactive waste tank sludges

    Energy Technology Data Exchange (ETDEWEB)

    KRUMHANSL,JAMES L.; BRADY,PATRICK V.; ZHANG,PENGCHU; ARTHUR,SARA E.; HUTCHERSON,SHEILA K.; LIU,J.; QIAN,M.; ANDERSON,HOWARD L.

    2000-05-19

    The US Department of Energy (DOE) has millions of gallons of high level nuclear waste stored in underground tanks at Hanford, Washington and Savannah River, South Carolina. These tanks will eventually be emptied and decommissioned. This will leave a residue of sludge adhering to the interior tank surfaces that may contaminate groundwaters with radionuclides and RCRA metals. Experimentation on such sludges is both dangerous and prohibitively expensive so there is a great advantage to developing artificial sludges. The US DOE Environmental Management Science Program (EMSP) has funded a program to investigate the feasibility of developing such materials. The following text reports on the success of this program, and suggests that much of the radioisotope inventory left in a tank will not move out into the surrounding environment. Ultimately, such studies may play a significant role in developing safe and cost effective tank closure strategies.

  14. Interim radiological safety standards and evaluation procedures for subseabed high-level waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Klett, R.D.

    1997-06-01

    The Seabed Disposal Project (SDP) was evaluating the technical feasibility of high-level nuclear waste disposal in deep ocean sediments. Working standards were needed for risk assessments, evaluation of alternative designs, sensitivity studies, and conceptual design guidelines. This report completes a three part program to develop radiological standards for the feasibility phase of the SDP. The characteristics of subseabed disposal and how they affect the selection of standards are discussed. General radiological protection standards are reviewed, along with some new methods, and a systematic approach to developing standards is presented. The selected interim radiological standards for the SDP and the reasons for their selection are given. These standards have no legal or regulatory status and will be replaced or modified by regulatory agencies if subseabed disposal is implemented. 56 refs., 29 figs., 15 tabs.

  15. Sample performance assessment of a high-level radioactive waste repository: sensitivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tkaczyk, A. [Iowa State Univ. of Science and Technology, Ames, IA (United States). Dept. of Mechanical Engineering

    2001-07-01

    The Yucca Mountain Project (YMP) is the USA's first attempt at long-term storage of High-Level Radioactive Waste (HLW). In theory, the reasoning for such a repository seems sound. In practice, there are many scenarios and cases to be considered while putting such a project into effect. Since a goal of YMP is to minimize dangers associated with long-term storage of HLW, it is important to estimate the dose rate to which current and future generations will be subjected. The lifetime of the repository is simulated to indicate the radiation dose rate to the maximally exposed individual; it is assumed that if the maximally exposed individual would not be harmed by the annual dose, the remaining population will be at even smaller risk. The determination of what levels of exposure can be deemed harmless is a concern, and the results from the simulations as compared against various regulations are discussed. (author)

  16. Characterization and Delivery of Hanford High-Level Radioactive Waste Slurry

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Michael G.; Denslow, Kayte M.; Lee, K. P.

    2014-11-15

    Two primary challenges to characterizing Hanford’s high-level radioactive waste slurry prior to transfer to a treatment facility are the ability to representatively sample million-gallon tanks and to estimate the critical velocity of the complex slurry. Washington River Protection Solutions has successfully demonstrated a sampling concept that minimizes sample errors by collecting multiple sample increments from a sample loop where the mixed tank contents are recirculated. Pacific Northwest National Laboratory has developed and demonstrated an ultrasonic-based Pulse-Echo detection device that is capable of detecting a stationary settled bed of solids in a pipe with flowing slurry. These two concepts are essential elements of a feed delivery strategy that drives the Hanford clean-up mission.

  17. The effect of high-level waste glass composition on spinel liquidus temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.; Riley, Brian J.; Crum, Jarrod V.; Matyas, Josef

    2014-01-15

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel solubility (c0) is a function of both glass composition and temperature (T). Previously reported models of TL as a function of composition are based on TL measured directly, which requires laborious experimental procedures. Viewing the curve of c0 versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates TL as a function of composition based on c0 data obtained with the X-ray diffraction technique.

  18. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Strum, M.J.; Weiss, H.; Farmer, J.C. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-06-01

    This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs.

  19. JNC thermodynamic database for performance assessment of high-level radioactive waste disposal system

    Energy Technology Data Exchange (ETDEWEB)

    Yui, Mikazu; Azuma, Jiro; Shibata, Masahiro [Japan Nuclear Cycle Development Inst., Tokai Works, Waste Isolation Research Division, Tokai, Ibaraki (Japan)

    1999-11-01

    This report is a summary of status, frozen datasets, and future tasks of the JNC (Japan Nuclear Cycle Development Institute) thermodynamic database (JNC-TDB) for assessing performance of high-level radioactive waste in geological environments. The JNC-TDB development was carried out after the first progress report on geological disposal research in Japan (H-3). In the development, thermodynamic data (equilibrium constants at 25degC, I=0) for important radioactive elements were selected/determined based on original experimental data using different models (e.g., SIT, Pitzer). As a result, the reliability and traceability of the data for most of the important elements were improved over those of the PNC-TDB used in H-3 report. For detailed information of data analysis and selections for each element, see the JNC technical reports listed in this document. (author)

  20. RADIOACTIVE HIGH LEVEL WASTE TANK PITTING PREDICTIONS: AN INVESTIGATION INTO CRITICAL SOLUTION CONCENTRATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E.

    2012-11-08

    A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel's susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-level wastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

  1. C-106 High-Level Waste Solids: Washing/Leaching and Solubility Versus Temperature Studies

    Energy Technology Data Exchange (ETDEWEB)

    GJ Lumetta; DJ Bates; PK Berry; JP Bramson; LP Darnell; OT Farmer III; LR Greenwood; FV Hoopes; RC Lettau; GF Piepel; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

    2000-01-26

    This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-level waste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

  2. The effect of high-level waste glass composition on spinel liquidus temperature

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Riley, Brian J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hrma, Pavel [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2012-11-15

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

  3. Mixing processes in high-level waste tanks. Progress report, September 15, 1996--September 14, 1997

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, P.F.

    1997-01-01

    'U.C. Berkeley has made excellent progress in the last year in building and running experiments and performing analysis to study mixing processes that can affect the distribution of fuel and oxygen in the air space of DOE high-level waste tanks, and the potential to create flammable concentrations at isolated locations, achieving all of the milestones outlined in the proposal. The DOE support has allowed the acquisition of key experimental equipment, and has funded the full-time efforts of one doctoral student and one postdoctoral researcher working on the project. In addition, one masters student and one other doctoral student, funded by external sources, have also contributed to the research effort. Flammable gases can be generated in DOE high-level waste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and ventilation systems, which remove the flammable gases and maintain a well-mixed condition in the tanks. Upon failure of the ventilation system, due to seismic or other events, however, it has proven more difficult to make strong arguments for well-mixed conditions, due to the potential for density-induced stratification which can potentially sequester fuel or oxidizer at concentrations significantly higher than average. This has complicated the task of defining the safety basis for tank operation. The author is currently developing numerical tools for modeling the transient evolution of fuel and oxygen concentrations in waste tanks following loss of ventilation. When used with reasonable grid resolutions, standard multi-dimensional fluid dynamics codes suffer from excessive numerical diffusion effects, which strongly over predict mixing and provide nonconservative estimates, particularly after stratification occurs. The National Institute of Standards and Technology (NIST) has developed

  4. Vapor Corrosion Response of Low Carbon Steel Exposed to Simulated High Level Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B

    2006-01-26

    A program to resolve the issues associated with potential vapor space corrosion and liquid/air interface corrosion in the Type III high level waste tanks is in place. The objective of the program is to develop understanding of vapor space (VSC) and liquid/air interface (LAIC) corrosion to ensure a defensible technical basis to provide accurate corrosion evaluations with regard to vapor space and liquid/air interface corrosion. The results of the FY05 experiments are presented here. The experiments are an extension of the previous research on the corrosion of tank steel exposed to simple solutions to corrosion of the steel when exposed to complex high level waste simulants. The testing suggested that decanting and the consequent residual species on the tank wall is the predominant source of surface chemistry on the tank wall. The laboratory testing has shown that at the boundary conditions of the chemistry control program for solutions greater than 1M NaNO{sub 3}{sup -}. Minor and isolated pitting is possible within crevices in the vapor space of the tanks that contain stagnant dilute solution for an extended period of time, specifically when residues are left on the tank wall during decanting. Liquid/air interfacial corrosion is possible in dilute stagnant solutions, particularly with high concentrations of chloride. The experimental results indicate that Tank 50 would be most susceptible to the potential for liquid/air interfacial corrosion or vapor space corrosion, with Tank 49 and 41 following, since these tanks are nearest to the chemistry control boundary conditions. The testing continues to show that the combination of well-inhibited solutions and mill-scale sufficiently protect against pitting in the Type III tanks.

  5. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

  6. Anticipated Degradation Modes of Metallic Engineered Barriers for High-Level Nuclear Waste Repositories

    Science.gov (United States)

    Rodríguez, Martín A.

    2014-03-01

    Metallic engineered barriers must provide a period of absolute containment to high-level radioactive waste in geological repositories. Candidate materials include copper alloys, carbon steels, stainless steels, nickel alloys, and titanium alloys. The national programs of nuclear waste management have to identify and assess the anticipated degradation modes of the selected materials in the corresponding repository environment, which evolves in time. Commonly assessed degradation modes include general corrosion, localized corrosion, stress-corrosion cracking, hydrogen-assisted cracking, and microbiologically influenced corrosion. Laboratory testing and modeling in metallurgical and environmental conditions of similar and higher aggressiveness than those expected in service conditions are used to evaluate the corrosion resistance of the materials. This review focuses on the anticipated degradation modes of the selected or reference materials as corrosion-resistant barriers in nuclear repositories. These degradation modes depend not only on the selected alloy but also on the near-field environment. The evolution of the near-field environment varies for saturated and unsaturated repositories considering backfilled and unbackfilled conditions. In saturated repositories, localized corrosion and stress-corrosion cracking may occur in the initial aerobic stage, while general corrosion and hydrogen-assisted cracking are the main degradation modes in the anaerobic stage. Unsaturated repositories would provide an oxidizing environment during the entire repository lifetime. Microbiologically influenced corrosion may be avoided or minimized by selecting an appropriate backfill material. Radiation effects are negligible provided that a thick-walled container or an inner shielding container is used.

  7. SPRAYED CLAY TECHNOLOGY FOR THE DEEP REPOSITORY OF HIGH-LEVEL RADIOACTIVE WASTE

    Directory of Open Access Journals (Sweden)

    Lucie Hausmannová

    2012-07-01

    Full Text Available The sealing barrier will play very important role in the Czech disposal concept of high level radioactive waste. It follows Swedish SKB3 design where granitic rock environment will host the repository. Swelling clay based materials as the most favorable for sealing purposes were selected. Such clays must fulfill certain requirements (e.g. on swelling properties, hydraulic conductivity or plasticity and must be stable for thousands of years. Better sealing behavior is obtained when the clay is compacted. Technology of the seal construction can vary according to its target dry density. Very high dry density is needed for buffer (sealing around entire canister with radioactive waste. Less strict requirements are on material backfilling the access galleries. It allows compaction to lower dry density than in case of buffer. One of potential technology for backfilling is to compact clay layers in most of the gallery profile by common compaction machines (rollers etc. and to spray clay into the uppermost part afterwards. The paper introduces the research works on sprayed clay technology performed at the Centre of Experimental Geotechnics of the Czech Technical University in Prague. Large scale in situ demonstration of filling of short drift in the Josef Gallery is also mentioned.

  8. PERFORMANCE OF A BURIED RADIOACTIVE HIGH LEVEL WASTE GLASS AFTER 24 YEARS

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C; Daniel Kaplan, D; Ned Bibler, N; David Peeler, D; John Plodinec, J

    2008-05-05

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

  9. Environment and oxidation state of molybdenum in simulated high level nuclear waste glass compositions

    Science.gov (United States)

    Short, R. J.; Hand, R. J.; Hyatt, N. C.; Möbus, G.

    2005-04-01

    Alkali borosilicate glasses containing between 20 and 35 wt% of a simulated high level nuclear waste stream with varying Li2O contents were melted under neutral (air) and reducing (nitrogen/hydrogen) conditions. XRD analysis of the as-cast glasses showed a tendency for the products to remain amorphous when melted under neutral conditions and for metallic silver to develop in the reduced melts. EXAFS analysis revealed (MoO4)2- tetrahedra in all glasses regardless of the sparge applied during melting. The glasses were heat treated to simulate an interruption to the cooling system used to prevent heat build-up in the vitrified product store. Powellite-type molybdate phases were found to develop in the heat treated samples and formed at lower waste loadings in glasses sparged with a reducing gas. A reduction in the quantity of Li2O lead to a reduction in the quantity of powellite-type molybdate phases. EDS showed the primary molybdate phase to be high in Sr and rare earth elements and TEM indicated that the presence of silver metal encouraged molybdate formation.

  10. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Simulant Studies

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jones, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-11-19

    Solubility testing with simulated High Level Waste tank heel solids has been conducted in order to evaluate two alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge washing efforts. Tests were conducted with non-radioactive pure phase metal reagents, binary mixtures of reagents, and a Savannah River Site PUREX heel simulant to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent and pure, dilute nitric acid toward dissolving the bulk non-radioactive waste components. A focus of this testing was on minimization of oxalic acid additions during tank cleaning. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid which is the current baseline chemical cleaning reagent. In a separate study, solubility tests were conducted with radioactive tank heel simulants using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species known to be drivers for Savannah River Site tank closure Performance Assessments. Permanganate-based cleaning methods were evaluated prior to and after oxalic acid contact.

  11. Chromium speciation and mobility in a high level nuclear waste vadose zone plume

    Science.gov (United States)

    Zachara, John M.; Ainsworth, Calvin C.; Brown, Gordon E.; Catalano, Jeffrey G.; McKinley, James P.; Qafoku, Odeta; Smith, Steven C.; Szecsody, James E.; Traina, Sam J.; Warner, Jeffrey A.

    2004-01-01

    Radioactive core samples containing elevated concentrations of Cr from a high level nuclear waste plume in the Hanford vadose zone were studied to asses the future mobility of Cr. Cr(VI) is an important subsurface contaminant at the Hanford Site. The plume originated in 1969 by leakage of self-boiling supernate from a tank containing REDOX process waste. The supernate contained high concentrations of alkali (NaOH ≈ 5.25 mol/L), salt (NaNO 3/NaNO 2 >10 mol/L), aluminate [Al(OH) 4- = 3.36 mol/L], Cr(VI) (0.413 mol/L), and 137Cs + (6.51 × 10 -5 mol/L). Water and acid extraction of the oxidized subsurface sediments indicated that a significant portion of the total Cr was associated with the solid phase. Mineralogic analyses, Cr valence speciation measurements by X-ray adsorption near edge structure (XANES) spectroscopy, and small column leaching studies were performed to identify the chemical retardation mechanism and leachability of Cr. While X-ray diffraction detected little mineralogic change to the sediments from waste reaction, scanning electron microscopy (SEM) showed that mineral particles within 5 m of the point of tank failure were coated with secondary, sodium aluminosilicate precipitates. The density of these precipitates decreased with distance from the source (e.g., beyond 10 m). The XANES and column studies demonstrated the reduction of 29-75% of the total Cr to insoluble Cr(III), and the apparent precipitation of up to 43% of the Cr(VI) as an unidentified, non-leachable phase. Both Cr(VI) reduction and Cr(VI) precipitation were greater in sediments closer to the leak source where significant mineral alteration was noted by SEM. These and other observations imply that basic mineral hydrolysis driven by large concentrations of OH - in the waste stream liberated Fe(II) from the otherwise oxidizing sediments that served as a reductant for CrO 42-. The coarse-textured Hanford sediments contain silt-sized mineral phases (biotite, clinochlore, magnetite, and

  12. Summary of Calcine Disposal Development Using Hot Isostatic Pressing

    Energy Technology Data Exchange (ETDEWEB)

    Bateman, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wahlquist, Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hart, Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States); McCartin, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Battelle Energy Alliance, LLC, has demonstrated the effectiveness of the hot isostatic press (HIP) process for treatment of hazardous high-level waste known as calcine that is stored at the Idaho Nuclear Technology and Engineering Center (INTEC) at Idaho National Laboratory. HIP trials performed with simulated calcines at Idaho National Laboratory’s Materials and Fuels Complex and an Australian Nuclear Science and Technology Organization facility from 2007 to 2010 produced a dense, monolithic waste form with increased chemical durability and effective (storage) volume reductions of ~10 to ~70% compared to granular calcine forms. In December 2009, the U.S. Department of Energy signed an amended Record of Decision selecting HIP technology as the treatment method for the 4,400 m3 of granular zirconia and alumina calcine stored at INTEC. Testing showed that HIP treatment reduces the risks associated with radioactive and hazardous constituent release, post-production handling, and long-term (repository) storage of calcines and would result in estimated storage cost savings in the billions of dollars. Battelle Energy Alliance has the ability to complete pilot-scale HIP processing of INTEC calcine, which is the next necessary step in implementing HIP processing as a calcine treatment method.

  13. Geomicrobiology of High Level Nuclear Waste-Contaminated Vadose Sediments at the Hanford Site, Washington State

    Energy Technology Data Exchange (ETDEWEB)

    Fredrickson, Jim K.; Zachara, John M.; Balkwill, David L.; Kennedy, David W.; Li, Shu-Mei W.; Kostandarithes, Heather M.; Daly, Michael J.; Romine, Margaret F.; Brockman, Fred J.

    2004-07-07

    Sediments from a high-level nuclear waste plume were collected as part of investigations to evaluate the potential fate and migration of contaminants in the subsurface. The plume originated from a leak that occurred in 1962 from a waste tank consisting of high concentrations of alkali, nitrate, aluminate, Cr(VI), 137Cs, and 99Tc. Investigations were initiated to determine the distribution of viable microorganisms in the vadose sediment samples, probe the phylogeny of cultivated and uncultivated members, and evaluate the ability of the cultivated organisms to survive acute doses of ionizing radiation. The populations of viable aerobic heterotrophic bacteria were generally low, from below detection to {approx}104 7 CFU g-1 but viable microorganisms were recovered from 11 of 16 samples including several of the most radioactive ones (e.g., > 10 ?Ci/g 137Cs). The isolates from the contaminated sediments and clone libraries from sediment DNA extracts were dominated by members related to known Gram-positive bacteria. Gram-positive bacteria most closely related to Arthrobacter species were the most common isolates among all samples but other high G+C phyla were also represented including Rhodococcus and Nocardia. Two isolates from the second most radioactive sample (>20 ?Ci 137Cs g-1) were closely related to Deinococcus radiodurans and were able to survive acute doses of ionizing radiation approaching 20kGy. Many of the Gram-positive isolates were resistant to lower levels of gamma radiation. These results demonstrate that Gram-positive bacteria, predominantly high G+C phyla, are indigenous to Hanford vadose sediments and some are effective at surviving the extreme physical and chemical stress associated with radioactive waste.

  14. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.L.; Breslin, J.J. (eds.)

    1981-01-01

    The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

  15. Implementation of industrial waste ferrochrome slag in conventional and low cement castables: Effect of calcined alumina

    Directory of Open Access Journals (Sweden)

    Pattem Hemanth Kumar

    2014-12-01

    Full Text Available A new class of conventional and low-cement ferrochrome slag-based castables were prepared from 40 wt.% ferrochrome slag and 45 wt.% calcined bauxite. Rest fraction varied between high alumina cement (HAC acting as hydraulic binder and calcined alumina as pore filling additive. Standard ASTM size briquettes were prepared for crushing and bending strengths evaluation, and the samples were then subjected to firing at 800, 1100 and 1300 °C for a soaking period of 3 h. The microstructure and refractory properties of the prepared castables have been investigated using X-ray diffraction (XRD, scanning electron microscopy (SEM, cold crushing strength, modulus of rupture and permanent linear changes (PLCs test. Castables show good volume stability (linear change <0.7% at 1300 °C. The outcomes of these investigations were efficacious and in accordance with previously reported data of similar compositions. High thermo-mechanical and physico-chemical properties were attained pointing out an outstanding potential to increase the refractory lining working life of non-recovery coke oven and reheating furnaces.

  16. Mercury removal at Idaho National Engineering and Environmental Laboratory's New Waste Calcining Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. C. Ashworth

    2000-02-27

    Technologies were investigated to determine viable processes for removing mercury from the calciner (NWCF) offgas system at the Idaho National Engineering and Environmental Laboratory. Technologies for gas phase and aqueous phase treatment were evaluated. The technologies determined are intended to meet EPA Maximum Achievable Control Technology (MACT) requirements under the Clean Air Act and Resource Conservation and Recovery Act (RCRA). Currently, mercury accumulation in the calciner off-gas scrubbing system is transferred to the tank farm. These transfers lead to accumulation in the liquid heels of the tanks. The principal objective for aqueous phase mercury removal is heel mercury reduction. The system presents a challenge to traditional methods because of the presence of nitrogen oxides in the gas phase and high nitric acid in the aqueous scrubbing solution. Many old and new technologies were evaluated including sorbents and absorption in the gas phase and ion exchange, membranes/sorption, galvanic methods, and UV reduction in the aqueous phase. Process modifications and feed pre-treatment were also evaluated. Various properties of mercury and its compounds were summarized and speciation was predicted based on thermodynamics. Three systems (process modification, NOxidizer combustor, and electrochemical aqueous phase treatment) and additional technology testing were recommended.

  17. Mercury Removal at Idaho National Engineering and Environmental Laboratory's New Waste Calcining Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ashworth, Samuel Clay; Wood, R. A.; Taylor, D. D.; Sieme, D. D.

    2000-03-01

    Technologies were investigated to determine viable processes for removing mercury from the calciner (NWCF) offgas system at the Idaho National Engineering and Environmental Laboratory. Technologies for gas phase and aqueous phase treatment were evaluated. The technologies determined are intended to meet EPA Maximum Achievable Control Technology (MACT) requirements under the Clean Air Act and Resource Conservation and Recovery Act (RCRA). Currently, mercury accumulation in the calciner off-gas scrubbing system is transferred to the tank farm. These transfers lead to accumulation in the liquid heels of the tanks. The principal objective for aqueous phase mercury removal is heel mercury reduction. The system presents a challenge to traditional methods because of the presence of nitrogen oxides in the gas phase and high nitric acid in the aqueous scrubbing solution. Many old and new technologies were evaluated including sorbents and absorption in the gas phase and ion exchange, membranes/sorption, galvanic methods, and UV reduction in the aqueous phase. Process modifications and feed pre-treatment were also evaluated. Various properties of mercury and its compounds were summarized and speciation was predicted based on thermodynamics. Three systems (process modification, NOxidizer combustor, and electrochemical aqueous phase treatment) and additional technology testing were recommended.

  18. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E.

    1996-09-01

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

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

    Energy Technology Data Exchange (ETDEWEB)

    1983-06-01

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

  20. 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

  1. Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.L. (ed.)

    1982-01-01

    The twenty papers in this volume are divided into three parts: site exploration and characterization; repository development and design; and waste package development and design. These papers represent the status of technology that existed in 1981 and 1982. Individual papers were processed for inclusion in the Energy Data Base.

  2. Evidence for dawsonite in Hanford high-level nuclear waste tanks.

    Science.gov (United States)

    Reynolds, Jacob G; Cooke, Gary A; Herting, Daniel L; Warrant, R Wade

    2012-03-30

    Gibbsite [Al(OH)(3)] and boehmite (AlOOH) have long been assumed to be the most prevalent aluminum-bearing minerals in Hanford high-level nuclear waste sludge. The present study shows that dawsonite [NaAl(OH)(2)CO(3)] is also a common aluminum-bearing phase in tanks containing high total inorganic carbon (TIC) concentrations and (relatively) low dissolved free hydroxide concentrations. Tank samples were probed for dawsonite by X-ray Diffraction (XRD), Scanning Electron Microscopy with Energy Dispersive Spectrometry (SEM-EDS) and Polarized Light Optical Microscopy. Dawsonite was conclusively identified in four of six tanks studied. In a fifth tank (AN-102), the dawsonite identification was less conclusive because it was only observed as a Na-Al bearing phase with SEM-EDS. Four of the five tank samples with dawsonite also had solid phase Na(2)CO(3) · H(2)O. The one tank without observable dawsonite (Tank C-103) had the lowest TIC content of any of the six tanks. The amount of TIC in Tank C-103 was insufficient to convert most of the aluminum to dawsonite (Al:TIC mol ratio of 20:1). The rest of the tank samples had much lower Al:TIC ratios (between 2:1 and 0.5:1) than Tank C-103. One tank (AZ-102) initially had dawsonite, but dawsonite was not observed in samples taken 15 months after NaOH was added to the tank surface. When NaOH was added to a laboratory sample of waste from Tank AZ-102, the ratio of aluminum to TIC in solution was consistent with the dissolution of dawsonite. The presence of dawsonite in these tanks is of significance because of the large amount of OH(-) consumed by dawsonite dissolution, an effect confirmed with AZ-102 samples. Copyright © 2012 Elsevier B.V. All rights reserved.

  3. Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

    KAUST Repository

    Salama, Amgad

    2015-11-01

    One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since the dawn of nuclear era. Albeit the relatively large number of research works that have been conducted to investigate temperature distribution surrounding waste canisters, they all abide to consider the host formations as homogeneous and isotropic. While this could be the case in some subsurface settings, in most cases, this is not true. In other words, subsurface formations are, in most cases, inherently anisotropic and heterogeneous. In this research, we show that even a slight difference in anisotropy of thermal conductivity of host rock with direction could have interesting effects on temperature fields. We investigate the effect of anisotropy angle (the angle the principal direction of anisotropy is making with the coordinate system) on the temperature field as well as on the maximum temperature attained in different barrier systems. This includes 0°, 30°, 45°, 60°, and 90°in addition to the isotropic case as a reference. We also consider the effect of anisotropy ratio (the ratio between the principal direction anisotropies) on the temperature fields and maximum temperature history. This includes ratios ranging between 1.5 and 4. Interesting patterns of temperature fields and profiles are obtained. It is found that the temperature contours are aligned more towards the principal direction of anisotropy. Furthermore the peak temperature in the buffer zone is found to be larger the smaller the anisotropy angle and vice versa. © 2015 Elsevier Ltd. All rights reserved.

  4. Low-temperature lithium diffusion in simulated high-level boroaluminosilicate nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Neeway, James J.; Kerisit, Sebastien N.; Gin, Stephane; Wang, Zhaoying; Zhu, Zihua; Ryan, Joseph V.

    2014-12-01

    Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solution of 6LiCl dissolved in dimethyl sulfoxide at 90 °C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, DLi, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D6Li ≈ 4.0-8.0 × 10-21 m2/s) exhibiting faster exchange than the more complex SON68 glass (DLi ≈ 2.0-4.0 × 10-21 m2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.

  5. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. (Lawrence Livermore National Lab., CA (USA)); Gdowski, G.E. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-06-01

    Three copper-based alloys, CDA 102 (oxygen-free, high-purity copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni), are candidates for the fabrication of high-level radioactive-waste disposal containers. Waste will include spent fuel assemblies from reactors as well as borosilicate glass, and will be sent to the prospective repository site at Yucca Mountain in Nye County, Nevada. The decay of radionuclides will result in the generation of substantial heat and in fluxes of gamma radiation outside the containers. In this environment, container materials might degrade by atmospheric oxidation, general aqueous phase corrosion, localized corrosion (LC), and stress corrosion cracking (SCC). This volume is a critical survey of available data on pitting and crevice corrosion of the copper-based candidates. Pitting and crevice corrosion are two of the most common forms of LC of these materials. Data on the SCC of these alloys is surveyed in Volume 4. Pitting usually occurs in water that contains low concentrations of bicarbonate and chloride anions, such as water from Well J-13 at the Nevada Test Site. Consequently, this mode of degradation might occur in the repository environment. Though few quantitative data on LC were found, a tentative ranking based on pitting corrosion, local dealloying, crevice corrosion, and biofouling is presented. CDA 102 performs well in the categories of pitting corrosion, local dealloying, and biofouling, but susceptibility to crevice corrosion diminishes its attractiveness as a candidate. The cupronickel alloy, CDA 715, probably has the best overall resistance to such localized forms of attack. 123 refs., 11 figs., 3 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-26

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

  7. Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

    Science.gov (United States)

    Pang, Bo; Becker, Frank

    2017-04-28

    Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  8. Sorption of strontium on uranyl peroxide: implications for a high-level nuclear waste repository.

    Science.gov (United States)

    Sureda, Rosa; Martínez-Lladó, Xavier; Rovira, Miquel; de Pablo, Joan; Casas, Ignasi; Giménez, Javier

    2010-09-15

    Strontium-90 is considered the most important radioactive isotope in the environment and one of the most frequently occurring radionuclides in groundwaters at nuclear facilities. The uranyl peroxide studtite (UO2O2 . 4H2O) has been observed to be formed in spent nuclear fuel leaching experiments and seems to have a relatively high sorption capacity for some radionuclides. In this work, the sorption of strontium onto studtite is studied as a function of time, strontium concentration in solution and pH. The main results obtained are (a) sorption is relatively fast although slower than for cesium; (b) strontium seems to be sorbed via a monolayer coverage of the studtite surface, (c) sorption has a strong dependence on ionic strength, is negligible at acidic pH, and increases at neutral to alkaline pH (almost 100% of the strontium in solution is sorbed above pH 10). These results point to uranium secondary solid phase formation on the spent nuclear fuel as an important mechanism for strontium retention in a high-level nuclear waste repository (HLNW). Copyright 2010 Elsevier B.V. All rights reserved.

  9. Biological ramifications of the subseabed disposal of high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, L.S.; Hessler, R.R.; Jackson, D.W.; Marietta, M.G.; Smith, K.L. Jr.; Talbert, D.M.; Yayanos, A.A.

    1980-01-01

    The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes, and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides.

  10. Thermal effects on clay rocks for deep disposal of high-level radioactive waste

    Directory of Open Access Journals (Sweden)

    Chun-Liang Zhang

    2017-06-01

    Full Text Available Thermal effects on the Callovo-Oxfordian and Opalinus clay rocks for hosting high-level radioactive waste were comprehensively investigated with laboratory and in situ experiments under repository relevant conditions: (1 stresses covering the range from the initial lithostatic state to redistributed levels after excavation, (2 hydraulic drained and undrained boundaries, and (3 heating from ambient temperature up to 90 °C–120 °C and a subsequent cooling phase. The laboratory experiments were performed on normal-sized and large hollow cylindrical samples in various respects of thermal expansion and contraction, thermally-induced pore water pressure, temperature influences on deformation and strength, thermal impacts on swelling, fracture sealing and permeability. The laboratory results obtained from the samples are consistent with the in situ observations during heating experiments in the underground research laboratories at Bure and Mont-Terri. Even though the claystones showed significant responses to thermal loading, no negative effects on their favorable barrier properties were observed.

  11. Geomechanical Studies on Granite Intrusions in Alxa Area for High-Level Radioactive Waste Disposal

    Directory of Open Access Journals (Sweden)

    Cheng Cheng

    2016-12-01

    Full Text Available Geological storage is an important concept for high-level radioactive waste (HLW disposal, and detailed studies are required to protect the environment from contamination by radionuclides. This paper presents a series of geomechanical studies on the site selection for HLW disposal in the Alxa area of China. Surface investigation in the field and RQD analyses on the drill cores are carried out to evaluate the rock mass quality. Laboratory uniaxial and triaxial compressive tests on the samples prepared from the drill cores are conducted to estimate the strength properties of the host rock. It is found that the NRG sub-area has massive granite intrusions, and NRG01 cored granite samples show the best rock quality and higher peak strength under various confinements (0–30 MPa. NRG01 granite samples are applied for more detailed laboratory studies considering the effects of strain rate and temperature. It is observed that the increasing strain rate from 1.0 × 10−5–0.6 × 10−2·s−1 can lead to a limited increase on peak strength, but a much more violent failure under uniaxial compressive tests on the NRG01 granite samples, and the temperature increasing from 20 °C–200 °C may result in a slight increase of UCS, as well as more ductile post-peak behavior in the triaxial compressive tests.

  12. Geochemistry research planning for the underground storage of high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Apps, J.A.

    1983-09-01

    This report is a preliminary attempt to plan a comprehensive program of geochemistry research aimed at resolving problems connected with the underground storage of high-level nuclear waste. The problems and research needs were identified in a companion report to this one. The research needs were taken as a point of departure and developed into a series of proposed projects with estimated manpowers and durations. The scope of the proposed research is based on consideration of an underground repository as a multiple barrier system. However, the program logic and organization reflect conventional strategies for resolving technological problems. The projects were scheduled and the duration of the program, critical path projects and distribution of manpower determined for both full and minimal programs. The proposed research was then compared with ongoing research within DOE, NRC and elsewhere to identify omissions in current research. Various options were considered for altering the scope of the program, and hence its cost and effectiveness. Finally, recommendations were made for dealing with omissions and uncertainties arising from program implementation. 11 references, 6 figures, 4 tables.

  13. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite.

    Science.gov (United States)

    Kaufhold, Stephan; Hassel, Achim Walter; Sanders, Daniel; Dohrmann, Reiner

    2015-03-21

    Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na-bentonites compared to the Ca-bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor. Copyright © 2014 Elsevier B.V. All rights reserved.

  14. Corrosion of container materials for disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Chun, K.S.; Park, H.S.; Yeon, J.W.; Ha, Y.K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    In the corrosion aspect of container for the deep geological disposal of high-level radioactive waste, disposal concepts and the related container materials, which have been developed by advanced countries, have been reviewed. The disposal circumstances could be divided into the saturated and the unsaturated zones. The candidate materials in the countries, which consider the disposal in the unsaturated zone, are the corrosion resistant materials such as supper alloys and stainless steels, but those in the saturated zone is cupper, one of the corrosion allowable materials. By the results of the pitting corrosion test of sensitized stainless steels (such as 304, 304L, 316 and 316L), pitting potential is decreased with the degree of sensitization and the pitting corrosion resistance of 316L is higher than others. And so, the long-term corrosion experiment with 316L stainless steel specimens, sebsitized and non-sensitized, under the compacted bentonite and synthetic granitic groundwater has been being carried out. The results from the experiment for 12 months indicate that no evidence of pitting corrosion of the specimens has been observed but the crevice corrosion has occurred on the sensitized specimens even for 3 months. (author). 33 refs., 19 figs., 10 tabs.

  15. Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff

    Energy Technology Data Exchange (ETDEWEB)

    Parsons, A.M.; Olague, N.E.; Gallegos, D.P. [Sandia National Labs., Albuquerque, NM (USA)

    1991-01-01

    Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information.

  16. Oxidative Alkaline leaching of Americium from simulated high-level nuclear waste sludges

    Energy Technology Data Exchange (ETDEWEB)

    Reed, Wendy A.; Garnov, Alexander Yu.; Rao, Linfeng; Nash, Kenneth L.; Bond, Andrew H.

    2004-01-23

    Oxidative alkaline leaching has been proposed to pre-treat the high-level nuclear waste sludges to remove some of the problematic (e.g., Cr) and/or non-radioactive (e.g., Na, Al) constituents before vitrification. It is critical to understand the behavior of actinides, americium and plutonium in particular, in oxidative alkaline leaching. We have studied the leaching behavior of americium from four different sludge simulants (BiPO{sub 4}, BiPO{sub 4 modified}, Redox, PUREX) using potassium permanganate and potassium persulfate in alkaline solutions. Up to 60% of americium sorbed onto the simulants is leached from the sludges by alkaline persulfate and permanganate. The percentage of americium leached increases with [NaOH] (between 1.0 and 5.0 M). The initial rate of americium leaching by potassium persulfate increases in the order BiPO{sub 4} sludge < Redox sludge < PUREX sludge. The data are most consistent with oxidation of Am{sup 3+} in the sludge to either AmO{sub 2}{sup +} or AmO{sub 2}{sup 2+} in solution. Though neither of these species is expected to exhibit long-term stability in solution, the potential for mobilization of americium from sludge samples would have to be accommodated in the design of any oxidative leaching process for real sludge samples.

  17. An approximate-reasoning-based method for screening high-level waste tanks for flammable gas

    Energy Technology Data Exchange (ETDEWEB)

    Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

    1998-07-01

    The in situ retention of flammable gas produced by radiolysis and thermal decomposition in high-level waste can pose a safety problem if the gases are released episodically into the dome space of a storage tank. Screening efforts at Hanford have been directed at identifying tanks in which this situation could exist. Problems encountered in screening motivated an effort to develop an improved screening methodology. Approximate reasoning (AR) is a formalism designed to emulate the kinds of complex judgments made by subject matter experts. It uses inductive logic structures to build a sequence of forward-chaining inferences about a subject. AR models incorporate natural language expressions known as linguistic variables to represent evidence. The use of fuzzy sets to represent these variables mathematically makes it practical to evaluate quantitative and qualitative information consistently. The authors performed a pilot study to investigate the utility of AR for flammable gas screening. They found that the effort to implement such a model was acceptable and that computational requirements were reasonable. The preliminary results showed that important judgments about the validity of observational data and the predictive power of models could be made. These results give new insights into the problems observed in previous screening efforts.

  18. A demonstration test of 4-group partitioning process with real high-level liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Morita, Y.; Yamaguchi, I.; Fujiwara, T.; Koizumi, H.; Tachimori, S. [Japan Atomic Energy Research Institute, Tokai-Mura, Ibaraki-Ken (Japan)

    2000-07-01

    The demonstration test of 4-Group Partitioning Process with concentrated real high-level liquid waste (HLLW) was carried out in the Partitioning Test Facility installed in a hot cell. More than 99.998% of Am and Cm were extracted from the HLLW with the organic solvent containing 0.5 M DIDPA - 0.1 M TBP, and more than 99.98% of Am and Cm were back-extracted with 4 M nitric acid. Np and Pu were extracted simultaneously, and more than 99.93% of Np and more than 99.98% of Pu were back-extracted with oxalic acid. In the denitration step for the separation of Tc and platinum group metals, more than 90% of Rh and more than 97% of Pd were precipitated. About half of Ru were remained in the de-nitrated solution, but the remaining Ru were quantitatively precipitated by neutralization of the de-nitrated solution to pH 6.7. In the adsorption step, both Sr and Cs were separated effectively. Decontamination factors for Cs and Sr were more than 10{sup 6} and 10{sup 4} respectively in all effluent samples. (authors)

  19. Strategy for addressing composition uncertainties in a Hanford high-level waste vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, M.F.; Piepel, G.F.

    1996-03-01

    Various requirements will be imposed on the feed material and glass produced by the high-level waste (HLW) vitrification plant at the Hanford Site. A statistical process/product control system will be used to control the melter feed composition and to check and document product quality. Two general types of uncertainty are important in HLW vitrification process/product control: model uncertainty and composition uncertainty. Model uncertainty is discussed by Hrma, Piepel, et al. (1994). Composition uncertainty includes the uncertainties inherent in estimates of feed composition and other process measurements. Because feed composition is a multivariate quantity, multivariate estimates of composition uncertainty (i.e., covariance matrices) are required. Three components of composition uncertainty will play a role in estimating and checking batch and glass attributes: batch-to-batch variability, within-batch uncertainty, and analytical uncertainty. This document reviews the techniques to be used in estimating and updating composition uncertainties and in combining these composition uncertainties with model uncertainty to yield estimates of (univariate) uncertainties associated with estimates of batch and glass properties.

  20. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, A.; Koploy, M.

    1991-12-01

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

  1. Attitudes and opposition in siting a high level nuclear waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeberg, L.; Viklund, M.; Truedsson, J

    1998-09-01

    In Sweden, the Swedish Nuclear Fuel and Waste Management Company (SKB) handles all issues concerning nuclear waste, including the siting process, in which the final outcome is intended to be a repository for high level nuclear waste placed deep down in bedrock. The main objective of the siting process is to find a host community fulfilling two important conditions: the safety demands have been met and agreements with the municipality can be accomplished. Only in such municipalities, so-called feasibility studies will be conducted. After conducting general studies in the whole country, SKB, in October 1992, sent letters with information about the intended feasibility studies to all Swedish municipalities. As a result, feasibility studies are or have been considered - and in some cases also been conducted - in eleven Swedish municipalities up until 1998. These are the municipalities where the attitudes and opposition towards a feasibility study, and possibly a final repository, are studied. The discussion can be divided into three main parts: Management of the siting process; Inherent `chaotic` processes and/or factors and risk perception. It is argued that two important problems could have been avoided at least partly: The citizens in many municipalities were uncertain of the relationship between a feasibility study and a final repository, and in many municipalities the citizens were afraid that the Government could overrule the municipal veto. Because of these fears, a common argument among the opponents of a feasibility study was: `to be sure of not receiving a final repository, we say no to a feasibility study`. Some inherent factors, more or less prevalent in the municipalities as well as in society in general, may also partly explain the outcome of the siting process. The municipalities in which the debate has been heated, and where public support has been more difficult to reach, share some common characteristics. Esp. in the municipalities in the north of

  2. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter. Preliminary settling and resuspension testing

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-05-01

    The full-scale, room-temperature Hanford Tank Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW) melter riser test system was successfully operated with silicone oil and magnetite particles at a loading of 0.1 vol %. Design and construction of the system and instrumentation, and the selection and preparation of simulant materials, are briefly reviewed. Three experiments were completed. A prototypic pour rate was maintained, based on the volumetric flow rate. Settling and accumulation of magnetite particles were observed at the bottom of the riser and along the bottom of the throat after each experiment. The height of the accumulated layer at the bottom of the riser, after the first pouring experiment, approximated the expected level given the solids loading of 0.1 vol %. More detailed observations of particle resuspension and settling were made during and after the third pouring experiment. The accumulated layer of particles at the bottom of the riser appeared to be unaffected after a pouring cycle of approximately 15 minutes at the prototypic flow rate. The accumulated layer of particles along the bottom of the throat was somewhat reduced after the same pouring cycle. Review of the time-lapse recording showed that some of the settling particles flow from the riser into the throat. This may result in a thicker than expected settled layer in the throat.

  3. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Actual Waste Testing with SRS Tank 5F Sludge

    Energy Technology Data Exchange (ETDEWEB)

    King, William D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, Michael S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Solubility testing with actual High Level Waste tank sludge has been conducted in order to evaluate several alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge sluicing efforts. Tests were conducted with archived Savannah River Site (SRS) radioactive sludge solids that had been retrieved from Tank 5F in order to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent toward dissolving the bulk non-radioactive waste components. Solubility tests were performed by direct sludge contact with the oxalic/nitric acid reagent and with sludge that had been pretreated and acidified with dilute nitric acid. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid following current baseline tank chemical cleaning methods. One goal of testing with the optimized reagent was to compare the total amounts of oxalic acid and water required for sludge dissolution using the baseline and optimized cleaning methods. A second objective was to compare the two methods with regard to the dissolution of actinide species known to be drivers for SRS tank closure Performance Assessments (PA). Additionally, solubility tests were conducted with Tank 5 sludge using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species.

  4. Characterization of high level nuclear waste glass samples following extended melter idling

    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)

    2015-06-16

    The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

  5. Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

    Energy Technology Data Exchange (ETDEWEB)

    Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

    2013-07-01

    Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

  6. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The scope of the Environmental Restoration and Waste Management (EM) Functional Area includes the programmatic controls associated with the management and operation of the Hanford Tank Farm Facility. The driving management organization implementing the programmatic controls is the Tank Farms Waste Management (WM)organization whose responsibilities are to ensure that performance objectives are established; and that measurable criteria for attaining objectives are defined and reflected in programs, policies and procedures. Objectives for the WM Program include waste minimization, establishment of effective waste segregation methods, waste treatment technology development, radioactive (low-level, high-level) hazardous and mixed waste transfer, treatment, and storage, applicability of a corrective action program, and management and applicability of a decontamination and decommissioning (D&D) program in future years.

  7. DEMONSTRATION AND EVALUATION OF POTENTIAL HIGH LEVEL WASTE MELTER DECONTAMINATION TECHNOLOGIES FOR SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Weger, Hans, Ph.D.; Kodanda, Raja Tilek Meruva; Mazumdar, Anindra; Srivastava, Rajiv Ph.D.; Ebadian, M.A. Ph.D.

    2003-02-27

    Four hand-held tools were tested for failed high-level waste melter decontamination and decommissioning (D&D). The forces felt by the tools during operation were measured using a tri-axial accelerometer since they will be operated by a remote manipulator. The efficiency of the tools was also recorded. Melter D&D consists of three parts: (1) glass fracturing: removing from the furnace the melted glass that can not be poured out through normal means, (2) glass cleaning: removing the thin layer of glass that has formed over the surface of the refractory material, and (3) K-3 refractory breakup: removing the K-3 refractory material. Surrogate glass, from a formula provided by the Savannah River Site, was melted in a furnace and poured into steel containers. K-3 refractory material, the same material used in the Defense Waste Processing Facility, was utilized for the demonstrations. Four K-3 blocks were heated at 1150 C for two weeks with a glass layer on top to simulate the hardened glass layer on the refractory surface in the melter. Tools chosen for the demonstrations were commonly used D&D tools, which have not been tested specifically for the different aspects of melter D&D. A jackhammer and a needle gun were tested for glass fracturing; a needle gun and a rotary grinder with a diamond face wheel (diamond grinder) were tested for glass cleaning; and a jackhammer, diamond grinder, and a circular saw with a diamond blade were tested for refractory breakup. The needle gun was not capable of removing or fracturing the surrogate glass. The diamond grinder only had a removal rate of 3.0 x 10-4 kg/s for K-3 refractory breakup and needed to be held firmly against the material. However, the diamond grinder was effective for glass cleaning, with a removal rate of 3.9 cm2/s. The jackhammer was successful in fracturing glass and breaking up the K-3 refractory block. The jackhammer had a glass-fracturing rate of 0.40 kg/s. The jackhammer split the K-3 refractory block into two

  8. Geochemical modelling of bentonite porewater in high-level waste repositories

    Science.gov (United States)

    Wersin, Paul

    2003-03-01

    The description of the geochemical properties of the bentonite backfill that serves as engineered barrier for nuclear repositories is a central issue for perfomance assessment since these play a large role in determining the fate of contaminants released from the waste. In this study the porewater chemistry of bentonite was assessed with a thermodynamic modelling approach that includes ion exchange, surface complexation and mineral equilibrium reactions. The focus was to identify the geochemical reactions controlling the major ion chemistry and acid-base properties and to explore parameter uncertainties specifically at high compaction degrees. First, the adequacy of the approach was tested with two distinct surface complexation models by describing recent experimental data performed at highly varying solid/liquid ratios and ionic strengths. The results indicate adequate prediction of the entire experimental data set. Second, the modelling was extended to repository conditions, taking as an example the current Swiss concept for high-level waste where the compacted bentonite backfill is surrounded by argillaceous rock. The main reactions controlling major ion chemistry were found to be calcite equilibrium and concurrent Na-Ca exchange reactions and de-protonation of functional surface groups. Third, a sensitivity analysis of the main model parameters was performed. The results thereof indicate a remarkable robustness of the model with regard to parameter uncertainties. The bentonite system is characterised by a large acid-base buffering capacity which leads to stable pH-conditions. The uncertainty in pH was found to be mainly induced by the pCO 2 of the surrounding host rock. The results of a simple diffusion-reaction model indicate only minor changes of porewater composition with time, which is primarily due to the geochemical similarities of the bentonite and the argillaceous host rock. Overall, the results show the usefulness of simple thermodynamic models to

  9. DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

    Directory of Open Access Journals (Sweden)

    HEUI-JOO CHOI

    2013-02-01

    Full Text Available Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  10. Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J. T. Case (DOE-ID); M. L. Renfro (INEEL)

    1998-12-01

    This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.

  11. Evaluation of blends bauxite-calcination-method red mud with other industrial wastes as a cementitious material: properties and hydration characteristics.

    Science.gov (United States)

    Zhang, Na; Liu, Xiaoming; Sun, Henghu; Li, Longtu

    2011-01-15

    Red mud is generated from alumina production, and its disposal is currently a worldwide problem. In China, large quantities of red mud derived from bauxite calcination method are being discharged annually, and its utilization has been an urgent topic. This experimental research was to evaluate the feasibility of blends red mud derived from bauxite calcination method with other industrial wastes for use as a cementitious material. The developed cementitious material containing 30% of the bauxite-calcination-method red mud possessed compressive strength properties at a level similar to normal Portland cement, in the range of 45.3-49.5 MPa. Best compressive strength values were demonstrated by the specimen RSFC2 containing 30% bauxite-calcination-method red mud, 21% blast-furnace slag, 10% fly ash, 30% clinker, 8% gypsum and 1% compound agent. The mechanical and physical properties confirm the usefulness of RSFC2. The hydration characteristics of RSFC2 were characterized by XRD, FTIR, (27)Al MAS-NMR and SEM. As predominant hydration products, ettringite and amorphous C-S-H gel are principally responsible for the strength development of RSFC2. Comparing with the traditional production for ordinary Portland cement, this green technology is easier to be implemented and energy saving. This paper provides a key solution to effectively utilize bauxite-calcination-method red mud. Copyright © 2010 Elsevier B.V. All rights reserved.

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

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

  14. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of FY2016 experiements

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

    Five experiments were completed with the full-scale, room temperature Hanford Waste Treatment and Immobilization Plant (WTP) high-level waste (HLW) melter riser test system to observe particle flow and settling in support of a crystal tolerant approach to melter operation. A prototypic pour rate was maintained based on the volumetric flow rate. Accumulation of particles was observed at the bottom of the riser and along the bottom of the throat after each experiment. Measurements of the accumulated layer thicknesses showed that the settled particles at the bottom of the riser did not vary in thickness during pouring cycles or idle periods. Some of the settled particles at the bottom of the throat were re-suspended during subsequent pouring cycles, and settled back to approximately the same thickness after each idle period. The cause of the consistency of the accumulated layer thicknesses is not year clear, but was hypothesized to be related to particle flow back to the feed tank. Additional experiments reinforced the observation of particle flow along a considerable portion of the throat during idle periods. Limitations of the system are noted in this report and may be addressed via future modifications. Follow-on experiments will be designed to evaluate the impact of pouring rate on particle re-suspension, the influence of feed tank agitation on particle accumulation, and the effect of changes in air lance positioning on the accumulation and re-suspension of particles at the bottom of the riser. A method for sampling the accumulated particles will be developed to support particle size distribution analyses. Thicker accumulated layers will be intentionally formed via direct addition of particles to select areas of the system to better understand the ability to continue pouring and re-suspend particles. Results from the room temperature system will be correlated with observations and data from the Research Scale Melter (RSM) at Pacific Northwest National Laboratory

  15. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

  16. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance.

  17. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

  18. Modelling magma-drift interaction at the proposed high-level radioactive waste repository at Yucca Mountain, Nevada, USA

    NARCIS (Netherlands)

    Woods, Andrew W.; Sparks, Steve; Bokhove, Onno; Lejeune, Anne-Marie; Connor, Charles B.; Hill, Britain E.

    2002-01-01

    We examine the possible ascent of alkali basalt magma containing 2 wt percent water through a dike and into a horizontal subsurface drift as part of a risk assessment for the proposed high-level radioactive waste repository beneath Yucca Mountain, Nevada, USA. On intersection of the dike with the

  19. Hanford Waste Vitrification Plant Quality Assurance Program description for high-level waste form development and qualification. Revision 3, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    The Hanford Waste Vitrification Plant Project has been established to convert the high-level radioactive waste associated with nuclear defense production at the Hanford Site into a waste form suitable for disposal in a deep geologic repository. The Hanford Waste Vitrification Plant will mix processed radioactive waste with borosilicate material, then heat the mixture to its melting point (vitrification) to forin a glass-like substance that traps the radionuclides in the glass matrix upon cooling. The Hanford Waste Vitrification Plant Quality Assurance Program has been established to support the mission of the Hanford Waste Vitrification Plant. This Quality Assurance Program Description has been written to document the Hanford Waste Vitrification Plant Quality Assurance Program.

  20. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Kaufhold, Stephan, E-mail: s.kaufhold@bgr.de [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); Hassel, Achim Walter [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Institute for Chemical Technology of Inorganic Materials, Johannes Kepler University Linz, Altenberger Straße 69, 4040 Linz (Austria); Sanders, Daniel [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Dohrmann, Reiner [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); LBEG, Landesamt für Bergbau, Energie und Geologie, Stilleweg 2, D-30655 Hannover (Germany)

    2015-03-21

    Graphical abstract: Corrosion at the bentonite iron interface proceeds unaerobically with formation of an 1:1 Fe silicate mineral. A series of exposure tests with different types of bentonites showed that Na–bentonites are slightly less corrosive than Ca–bentonites and highly charges smectites are less corrosive compared to low charged ones. The formation of a patina was observed in some cases and has to be investigated further. - Highlights: • At the iron bentonite interface a 1:1 Fe layer silicate forms upon corrosion. • A series of iron–bentonite corrosion products showed slightly less corrosion for Na-rich and high-charged bentonites. • In some tests the formation of a patina was observed consisting of Fe–silicate, which has to be investigated further. - Abstract: Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na–bentonites compared to the Ca–bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe

  1. Multilayer Protective Coatings for High-Level Nuclear Waste Storage Containers

    Science.gov (United States)

    Fusco, Michael

    Corrosion-based failures of high-level nuclear waste (HLW) storage containers are potentially hazardous due to a possible release of radionuclides through cracks in the canister due to corrosion, especially for above-ground storage (i.e. dry casks). Protective coatings have been proposed to combat these premature failures, which include stress-corrosion cracking and hydrogen-diffusion cracking, among others. The coatings are to be deposited in multiple thin layers as thin films on the outer surface of the stainless steel waste basket canister. Coating materials include: TiN, ZrO2, TiO2, Al 2O3, and MoS2, which together may provide increased resistances to corrosion and mechanical wear, as well as act as a barrier to hydrogen diffusion. The focus of this research is on the corrosion resistance and characterization of single layer coatings to determine the possible benefit from the use of the proposed coating materials. Experimental methods involve electrochemical polarization, both DC and AC techniques, and corrosion in circulating salt brines of varying pH. DC polarization allows for estimation of corrosion rates, passivation behavior, and a qualitative survey of localized corrosion, whereas AC electrochemistry has the benefit of revealing information about kinetics and interfacial reactions that is not obtainable using DC techniques. Circulation in salt brines for nearly 150 days revealed sustained adhesion of the coatings and minimal weight change of the steel samples. One-inch diameter steel coupons composed of stainless steel types 304 and 316 and A36 low alloy carbon steel were coated with single layers using magnetron sputtering with compound targets in an inert argon atmosphere. This resulted in very thin films for the metal-oxides based on low sputter rates. DC polarization showed that corrosion rates were very similar between bare and coated stainless steel samples, whereas a statistically significant decrease in uniform corrosion was measured on coated

  2. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  3. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  4. Confidence improvement of disosal safety bydevelopement of a safety case for high-level radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Min Hoon; Ko, Nak Youl; Jeong, Jong Tae; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

  5. Analogues to features and processes of a high-level radioactive waste repository proposed for Yucca Mountain, Nevada

    Science.gov (United States)

    Simmons, Ardyth M.; Stuckless, John S.; with a Foreword by Abraham Van Luik, U.S. Department of Energy

    2010-01-01

    Natural analogues are defined for this report as naturally occurring or anthropogenic systems in which processes similar to those expected to occur in a nuclear waste repository are thought to have taken place over time periods of decades to millennia and on spatial scales as much as tens of kilometers. Analogues provide an important temporal and spatial dimension that cannot be tested by laboratory or field-scale experiments. Analogues provide one of the multiple lines of evidence intended to increase confidence in the safe geologic disposal of high-level radioactive waste. Although the work in this report was completed specifically for Yucca Mountain, Nevada, as the proposed geologic repository for high-level radioactive waste under the U.S. Nuclear Waste Policy Act, the applicability of the science, analyses, and interpretations is not limited to a specific site. Natural and anthropogenic analogues have provided and can continue to provide value in understanding features and processes of importance across a wide variety of topics in addressing the challenges of geologic isolation of radioactive waste and also as a contribution to scientific investigations unrelated to waste disposal. Isolation of radioactive waste at a mined geologic repository would be through a combination of natural features and engineered barriers. In this report we examine analogues to many of the various components of the Yucca Mountain system, including the preservation of materials in unsaturated environments, flow of water through unsaturated volcanic tuff, seepage into repository drifts, repository drift stability, stability and alteration of waste forms and components of the engineered barrier system, and transport of radionuclides through unsaturated and saturated rock zones.

  6. Properties of the platinoid fission products during vitrification of high-level radioactive waste

    Science.gov (United States)

    Gong, W.; Lutze, W.; Perez-Cardenas, F.; Matlack, K. S.; Pegg, I. L.

    2006-05-01

    Platinoid fission products present in high-level nuclear wastes present particular challenges to their treatment by vitrification. The platinoid metals Ru, Rh, Pd, and their compounds are sparingly soluble in borosilicate glass melts. During glass melting under oxidizing conditions, the platinoids form small crystals of highly dense solid intermetallic phases and oxides. Under reducing conditions, the platinoids form only intermetallic phases. A fraction of these crystals settles to the bottom of the melting furnace, forming an immobile sludge. The fraction settling reported in the literature is highly variable. In the present work, the fraction settling was found to be >90% under reducing conditions but only 10 to 20% under oxidizing conditions. The thickness of the sludge layer depends on the volume fraction of platinoid crystals in the sludge, which is poorly known (typically ~0.06 under oxidizing conditions). Since the electrical conductivity of the sludge can be >10X that of the melt, in joule-heated melters the presence of such a layer can lead to diversion of the electric current, thereby compromising melter operability. The time to failure by this mechanism is clearly of practical importance. A variety of data are required in order to estimate the time to failure due to this mechanism and such data must be obtained under conditions representative of those in a full-size melting furnace. We have acquired such data using a melting furnace installed in our laboratory. This furnace is a one-third scale prototype of the system to be used for the vitrification of defense HLW at Hanford, WA. In the present work, simulated Hanford HLW material was combined with glass formers to produce a melter feed slurry that was then spiked with the platinoids. Over one thousand chemical and optical analyses were performed on hundreds of samples taken from the feed, various locations inside the furnace, the glass melt during pouring, the solid glass, and various locations along

  7. Reducing the likelihood of future human activities that could affect geologic high-level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    1984-05-01

    The disposal of radioactive wastes in deep geologic formations provides a means of isolating the waste from people until the radioactivity has decayed to safe levels. However, isolating people from the wastes is a different problem, since we do not know what the future condition of society will be. The Human Interference Task Force was convened by the US Department of Energy to determine whether reasonable means exist (or could be developed) to reduce the likelihood of future human unintentionally intruding on radioactive waste isolation systems. The task force concluded that significant reductions in the likelihood of human interference could be achieved, for perhaps thousands of years into the future, if appropriate steps are taken to communicate the existence of the repository. Consequently, for two years the task force directed most of its study toward the area of long-term communication. Methods are discussed for achieving long-term communication by using permanent markers and widely disseminated records, with various steps taken to provide multiple levels of protection against loss, destruction, and major language/societal changes. Also developed is the concept of a universal symbol to denote Caution - Biohazardous Waste Buried Here. If used for the thousands of non-radioactive biohazardous waste sites in this country alone, a symbol could transcend generations and language changes, thereby vastly improving the likelihood of successful isolation of all buried biohazardous wastes.

  8. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

    2013-01-17

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

  9. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

    2013-07-01

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

  10. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  11. Too hot to touch: the problem of high-level nuclear waste

    National Research Council Canada - National Science Library

    Alley, William M; Alley, Rosemarie

    2012-01-01

    ... Mountain repository project. William and Rosemarie Alley provide an engaging and authoritative account of the controversies and possibilities surrounding disposal of nuclear waste in the US, with reference also to other countries around the world...

  12. Derived Requirements for Double Shell Tank (DST) High Level Waste (HLW) Auxiliary Solids Mobilization

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, A.R.

    2000-02-28

    The potential need for auxiliary double-shell tank waste mixing and solids mobilization requires an evaluation of optional technologies. This document formalizes those operating and design requirements needed for further engineering evaluations.

  13. General and Localized Corrosion of Outer Barrier of High-Level Waste Container in Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.; McCright, D.; Gdowski, G.; Wang, F.; Summers, T.; Bedrossian, P.; Horn, J.; Lian, T.; Estill, J.; Lingenfelter, A.; Halsey, W.

    2000-05-02

    As described in the License Application Design Selection Report, the recommended waste, package design is Engineering Design Alternative II (CRWMS M&O 1999). This design includes a double-wall waste package (WP) underneath a protective drip shield (DS). purpose and scope of the process-level model described here is to account for both general and localized corrosion of the waste package outer barrier (WPOB), which assumed to be Alloy 22 (UNS N06022-21Cr-13Mo-4Fe-3W-2C-Ni) (ASTM 1997a). This model will include several sub-models, which will account for dry oxidation (DOX), humid air corrosion (HAC), general corrosion (GC) in the aqueous phase, and localized corrosion (LC) the aqueous phase. This model serves as a feed to the waste package degradation (WAPDEG) code for performance, assessment.

  14. High-level waste-basalt interactions. Annual progress report, February 1, 1977--September 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.; Scheetz, B.E.

    1978-05-01

    Commercial radioactive waste can be placed under ground in a basalt repository to contain significant amounts of radioactive decay heat for the first hundred or so years, which constitutes the ''thermal period'' of waste isolation, if the feasibility is determined that a basalt geology is a suitable medium for storage of radioactive wastes. Several physical-chemical changes analogous to natural geochemical processes can occur in and around this repository during the thermal period. The waste canister can act as a heat source and cause changes in the mineralogy and properties of the surrounding basalts. Geochemically, this is ''contact metamorphism.'' This phenomenon needs to be investigated because it could affect the behavior of the basalt with regard to migration of long-lived radionuclides away from the immediate repository. It is well known that even the relatively low-grade hydrothermal conditions possible in the repository (temperatures up to 400 degrees Centigrade; pressures up to 300 bars) can cause extensive modifications in rocks and minerals. At the end of the thermal period, the residue of the original waste plus the waste-basalt interaction products would constitute the actual waste form (or ''source term'') subject to the low-temperature leaching and migration processes under investigation in other laboratories. During the last eight months of fiscal year 1977, a program was initiated at The Pennsylvania State University which had as its objective the determination of the nature and implication of any chemical or mineralogical changes in, or interactions between, each candidate radioactive waste form and representative Columbia River Basalt under the various relevant repository conditions during the thermal period. Results of these investigations are given.

  15. Survey of matrix materials for solidified radioactive high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Gurwell, W.E.

    1981-09-01

    Pacific Northwest Laboratory (PNL) has been investigating advanced waste forms, including matrix waste forms, that may provide a very high degree of stability under the most severe repository conditions. The purpose of this study was to recommend practical matrix materials for future development that most enhance the stability of the matrix waste forms. The functions of the matrix were reviewed. Desirable matrix material properties were discussed and listed relative to the matrix functions. Potential matrix materials were discussed and recommendations were made for future matrix development. The matrix mechanically contains waste cores, reduces waste form temperatures, and is capable of providing a high-quality barrier to leach waters. High-quality barrier matrices that separate and individually encapsulate the waste cores are fabricated by powder fabrication methods, such as sintering, hot pressing, and hot isostatic pressing. Viable barrier materials are impermeable, extremely corrosion resistant, and mechanically strong. Three material classes potentially satisfy the requirements for a barrier matrix and are recommended for development: titanium, glass, and graphite. Polymers appear to be marginally adequate, and a more thorough engineering assessment of their potential should be made.

  16. Building the institutional capacity for managing commercial high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-05-01

    In July 1981, the Office of Nuclear Waste Management of the Department of Energy contracted with the National Academy of Public Administration for a study of institutional issues associated with the commercial radioactive waste management program. The two major sets of issues which the Academy was asked to investigate were (1) intergovernmental relationships, how federal, state, local and Indian tribal council governments relate to each other in the planning and implementation of a waste management program, and (2) interagency relationships, how the federal agencies with major responsibilities in this public policy arena interact with each other. The objective of the study was to apply the perspectives of public administration to a difficult and controversial question - how to devise and execute an effective waste management program workable within the constraints of the federal system. To carry out this task, the Academy appointed a panel composed of individuals whose background and experience would provide the several types of knowledge essential to the effort. The findings of this panel are presented along with the executive summary. The report consists of a discussion of the search for a radioactive waste management strategy, and an analysis of the two major groups of institutional issues: (1) intergovernmental, the relationship between the three major levels of government; and (2) interagency, the relationships between the major federal agencies having responsibility for the waste management program.

  17. STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

    Directory of Open Access Journals (Sweden)

    Trpimir Kujundžić

    2012-07-01

    Full Text Available Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter deposition holes. Drilling technology of the deposition holes depends on repository concept and geological and geomechanical characteristics of the rock. The deposition holes are mechanically excavated since drill & blast is not a possible method due to requirements on final geometry like surface roughness etc. Different methods of drilling large diameter boreholes for deposition of high-level waste and spent nuclear fuel are described. Comparison of methods is made considering performance and particularities in technology.

  18. STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

    OpenAIRE

    Trpimir Kujundžić; Tomislav Korman; Marija Macenić

    2012-01-01

    Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter...

  19. Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

    1996-02-01

    High-level radioactive waste stored in tanks at the U.S. Department of Energy`s (DOE`s) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H{sub 2} and NH{sub 3} during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H{sub 2} and NH{sub 3}. Both laboratory-scale and pilot-scale studies at SRTC have documented the H{sub 2} and NH{sub 3} generation phenomenal Because H{sub 2} and NH{sub 3} may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H{sub 2} generation rate and the NH{sub 3} generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste.

  20. PNL vitrification technology development project high-waste loaded high-level waste glasses for high-temperature melter: Letter report

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.; Hrma, P.R.

    1996-02-01

    For vitrification of high-level wastes (HLW) at the Hanford Site, a Joule-heated overflow type melter with bottom draining capability and capable of operating at temperatures up to 1500{degrees}C is being developed. The original proposed Hanford Waste Vitrification Plant (HWVP) melter used a 1150{degrees}C processing temperature and was tested using glasses with up to 28 wt% waste oxide loading for NCAW (Neutralized Current Acid Waste). The goal of the high-temperature melter (HTM) is the volume reduction of the final product and increase of the waste processing rate by processing high-waste loaded glasses at higher temperatures. This would dramatically decrease waste disposal and processing costs. The aim of glass development for the HTM is to determine compositions and melting temperatures for processible and acceptable glasses with a high waste loading. Glass property/composition models for viscosity and liquidus temperature developed in the Glass Envelope Definition (GED) study were used. The results of glass formulation and experimental testing are presented for NCAW and DST/SST (Double-Shell Tank/Single-Shell Tank) Blend waste. Although the purpose of this report was to summarize the glass development study with Blend waste only, the results with NCAW were needed because glass development with Blend waste was based on the results from the glass development study with NCAW.

  1. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    Energy Technology Data Exchange (ETDEWEB)

    F. Hua; P. Pasupathi; N. Brown; K. Mon

    2005-09-19

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

  2. Development of a Universal Canister for Disposal of High-Level Waste in Deep Boreholes.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gomberg, Steve [USDOE, Washington, DC (United States)

    2015-11-01

    The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock. In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfer, transportation, and disposal) could facilitate the eventual disposal of these wastes. Development of specifications for the universal canister system will consider the regulatory requirements that apply to storage, transportation, and disposal of the capsules, as well as operational requirements and limits that could affect the design of the canister (e.g., deep borehole diameter). In addition, there are risks and technical challenges that need to be recognized and addressed as Universal Canister system specifications are developed. This paper provides an approach to developing specifications for such a canister system that is integrated with the overall efforts of the DOE’s Used Fuel Disposition Campaign's Deep Borehole Field Test and compatible with planned storage of potential borehole-candidate wastes.

  3. Numerical simulation of high-level radioactive nuclear waste glass production

    Energy Technology Data Exchange (ETDEWEB)

    Choi, I.G. [Westinghouse Savannah River Co., Aiken, SC (United States); Ungan, A. [Purdue Univ., Indianapolis, IN (United States). Dept. of Mechanical Engineering

    1991-12-31

    Vitrification of radioactive waste has become an international approach for converting highly radioactive wastes into a durable solid prior to placing them in a permanent disposal repository. The technology for the process is not new. The conversion melter is a direct descendant of all electric melters used for manufacturing of some commercial glass types. Therefore, the vitrification process of radioactive wastes inherits typical problems of all electric furnaces and creates some other specific problems such as noble metal sedimentation. The noble metals and nickel sulfides in the melter are heavier than molten glass and have a low solubility. In a reducing condition, these metals amalgamate and tend to settle on the melter floor. The metal deposit resulting from this settling has a potential to short circuit the melter. The objective of this paper is to identify the typical problems that have been encountered in the waste melter operations and to address how these problems can be tackled using state-of-the-art numerical simulation techniques. It is believed that the large amount of pilot-scale melter experience throughout the world, combined with the knowledge gained from state-of-the-art computer modeling techniques would give assurance that the existing and future radioactive wastes can be effectively converted into a durable glass material and safely placed in a permanent repository.

  4. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    Energy Technology Data Exchange (ETDEWEB)

    PACQUET, E.A.

    2000-07-20

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

  5. 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

  6. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

  7. Noble metal behavior during melting of simulated high-level nuclear waste glass feeds

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1993-04-01

    Noble metals and their oxides can settle in waste glass melters and cause electrical shorting. Simulated waste feeds from Hanford, Savannah River, and Germany were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600{degrees}C--1000{degrees}C and examined by electron microscopy to determine shapes, sizes, and distribution of noble metal particles as a function of temperature. Individual noble metal particles and agglomerates of rhodium (Rh), ruthenium (RuO{sub 2}), and palladium (Pd), as well as their alloys, were seen. the majority of particles and agglomerates were generally less than 10 microns; however, large agglomerations (up to 1 mm) were found in the German feed. Detailed particle distribution and characterization was performed for a Hanford waste to provide input to computer modeling of particle settling in the melter.

  8. Evaluation the microwave heating of spinel crystals in high-level waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Washington, A. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL)

    2015-08-18

    In this report, the microwave heating of a crystal-free and a partially (24 wt%) trevorite-crystallized waste glass simulant were evaluated. The results show that a 500 mg piece of partially crystallized waste glass can be heated from room-temperature to above 1600 °C (as measured by infrared radiometry) within 2 minutes using a single mode, highly focused, 2.45 GHz microwave, operating at 300 W. X-ray diffraction measurements show that the partially crystallized glass experiences an 87 % reduction in trevorite following irradiation and thermal quenching. When a crystal-free analogue of the same waste glass simulant composition is exposed to the same microwave radiation it could not be heated above 450 °C regardless of the heating time.

  9. The extraction of rare earth elements from ICPP sodium-bearing waste and dissolved zirconium calcine by CMP and TRUEX solvents

    Energy Technology Data Exchange (ETDEWEB)

    Todd, T.A.; Glagolenko, I.Y.; Herbst, R.S.; Brewer, K.N.

    1995-11-01

    The extraction of stable isotopes of Eu and Ce was investigated from simulated sodium-bearing waste (SBW) and dissolved zirconium calcine by TRUEX and CMP solvents at the Idaho Chemical Processing Plant (ICPP). Single batch contacts were carried out in order to evaluate the rare earth behavior in the extraction, scrub, strip and wash sections for the proposed flowsheets. It has been shown that these lanthanides are efficiently extracted from the sodium-bearing wastes into either solvent, are not scrubbed and are stripped from both of the extractants with dilute HEDPA. The extraction distribution coefficients for Ce and Eu are higher in the TRUEX solvent (D{sub Ce} = 11.7, D{sub Eu} = 14.9) compared with CMP (D{sub Ce} = 9.3, D{sub Eu} = 7.23) for SBW. The extraction distribution coefficients for Ce and Eu are considerably less in the TRUEX solvent (D{sub Ce}=1.13, D{sub Eu}=1.8) than in the CMP solvent (D{sub Ce}=7.4, D{sub Eu=}6.1) for dissolved zirconium calcine feeds. The lower distribution coefficients for the extraction of lanthanides in the TRUEX/dissolved zirconium calcine system can be explained by zirconium loading of the solvent. The data obtained also confirmed that Ce and Eu can be used as non-radioactive surrogates for Am in separation experiments with acidic solutions.

  10. A review on immobilization of phosphate containing high level nuclear wastes within glass matrix--present status and future challenges.

    Science.gov (United States)

    Sengupta, Pranesh

    2012-10-15

    Immobilization of phosphate containing high level nuclear wastes within commonly used silicate glasses is difficult due to restricted solubility of P(2)O(5) within such melts and its tendency to promote crystallization. The situation becomes more adverse when sulfate, chromate, etc. are also present within the waste. To solve this problem waste developers have carried out significant laboratory scale research works in various phosphate based glass systems and successfully identified few formulations which apparently look very promising as they are chemically durable, thermally stable and can be processed at moderate temperatures. However, in the absence of required plant scale manufacturing experiences it is not possible to replace existing silicate based vitrification processes by the phosphate based ones. A review on phosphate glass based wasteforms is presented here. Copyright © 2012 Elsevier B.V. All rights reserved.

  11. DEVELOPMENT OF AN IMPROVED SODIUM TITANATE FOR THE PRETREATMENT OF HIGH LEVEL NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Hobbs, D

    2007-11-15

    High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90 and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST) and caustic side solvent extraction, for {sup 137}Cs removal. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239 and Pu-240. This paper describes recent results to produce an improved sodium titanate material that exhibits increased removal kinetics and capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material.

  12. Accelerator-driven transmutation of high-level waste from the defense and commercial sectors

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, C.; Arthur, E.; Beard, C. [and others

    1996-09-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The major goal has been to develop accelerator transmutation of waste (ATW) system designs that will thoroughly and rapidly transmute nuclear waste, including plutonium from dismantled weapons and spent reactor fuel, while generating useful electrical power and without producing a long-lived radioactive waste stream. We have identified and quantified the unique qualities of subcritical nuclear systems and their capabilities in bringing about the complete destruction of plutonium. Although the 1191 subcritical systems involved in our most effective designs radically depart from traditional nuclear reactor concepts, they are based on extrapolations of existing technologies. Overall, care was taken to retain the highly desired features that nuclear technology has developed over the years within a conservative design envelope. We believe that the ATW systems designed in this project will enable almost complete destruction of nuclear waste (conversion to stable species) at a faster rate and without many of the safety concerns associated with the possible reactor approaches.

  13. The apparent solubility of aluminum (III) in Hanford high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, Jacob G.

    2012-12-01

    The solubility of aluminum in Hanford nuclear waste impacts on the process ability of the waste by a number of proposed treatment options. For many years, Hanford staff has anecdotally noted that aluminum appears to be considerably more soluble in Hanford waste than the simpler electrolyte solutions used as analogues. There has been minimal scientific study to confirm these anecdotal observations, however. The present study determines the apparent solubility product for gibbsite in 50 tank samples. The ratio of hydroxide to aluminum in the liquid phase for the samples is calculated and plotted as a function of total sodium molarity. Total sodium molarity is used as a surrogate for ionic strength, because the relative ratios of mono, di and trivalent anions are not available for all of the samples. These results were compared to the simple NaOH-NaAl(OH{sub 4})H{sub 2}O system, and the NaOH-NaAl(OH{sub 4})NaCl-H{sub 2}O system data retrieved from the literature. The results show that gibbsite is apparently more soluble in the samples than in the simple systems whenever the sodium molarity is greater than two. This apparent enhanced solubility cannot be explained solely by differences in ionic strength. The change in solubility with ionic strength in simple systems is small compared to the difference between aluminum solubility in Hanford waste and the simple systems. The reason for the apparent enhanced solubility is unknown, but could include. kinetic or thermodynamic factors that are not present in the simple electrolyte systems. Any kinetic explanation would have to explain why the samples are always supersaturated whenever the sodium molarity is above two. Real waste characterization data should not be used to validate thermodynamic solubility models until it can be confirmed that the apparent enhanced gibbsite solubility is a thermodynamic effect and not a kinetic effect.

  14. Complex geological investigations to select a site for high-level waste disposal in the Krasnoyarsk region (Russia)

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, E.B.; Savonenkov, V.G.; Shabalev, S.I.; Rogozin, Yu.M. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation); Lyubtseva, E.F. [St. Petersburg State University, St. Petersburg (Russian Federation); Revenko, Yu.A.; Sabaev, Yu.I.; Nosukhin, A.V. [Chemical Mining Company, Krasnoyarsk (Russian Federation); Milovidov, V.L. [VNIPI Promtechnologii, Moscow (Russian Federation); Lukina, N.V. [GIN RAN, Moscow (Russian Federation); Lopatin, A.P. [State Centre ' Priroda' , Krasnoyarsk branch, Krasnoyarsk (Russian Federation); Datsenko, V.M. [KNIIGiMS, Krasnoyarsk (Russian Federation); Kryzhanovsky, V.A. [' Krasnoyarskgeolsyomka' , Krasnoyarsk (Russian Federation)

    1998-07-01

    The disposal of high-level radioactive waste (HLW) in deep geological formations is now considered by the international community as the ultimate stage in the optimum solution to the general problem of radioactive waste management, although other, more hypothetical, possibilities, (e.g. transmutation, space disposal) have been considered. lt is no exaggeration to state that the creation of an underground HLW repository in the Krasnoyarsk region is the greatest of the key environmental and economic problems governing the very existence of the large nuclear company, the Mining and Chemical Company (MCC), which generates considerable quantities of radioactive wastes, including HLW. On the other hand, radioactive waste disposal is not merely a problem for the Krasnoyarsk region, but is one of the key problems determining the future development of the Russian nuclear power industry in general, particularly if the nuclear fuel cycle includes the reprocessing of spent nuclear fuel which is planned for the RT-2 plant currently under construction in the Krasnoyarsk region. lt is no accident, therefore, that the Russian Federal programme on radioactive waste management includes a special topic devoted to characterisation work in the Krasnoyarsk region with a view to selecting a site for the construction of a repository. The results of a number of international scientific research programmes have demonstrated that in the case of long-term storage of HLW and intermediate-level wastes (ILW) in the earth's surface, the risk to the population and to the biosphere in general is too great and, consequently, unacceptable. Thus, the concept of an underground radioactive waste repository has wide support among the leading nuclear countries, including the U.K., France, U.S.A., Sweden, Finland and Germany. In these countries, projects to construct repositories for the geological disposal of radioactive waste are either underway or are planned. 1 ref.

  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. Probability, consequences, and mitigation for lightning strikes of Hanford high level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Zach, J.J.

    1996-06-05

    The purpose of this report is to summarize selected lightning issues concerning the Hanford Waste Tanks. These issues include the probability of a lightning discharge striking the area immediately adjacent to a tank including a riser, the consequences of significant energy deposition from a lightning strike in a tank, and mitigating actions that have been or are being taken. The major conclusion of this report is that the probability of a lightning strike deposition sufficient energy in a tank to cause an effect on employees or the public is unlikely;but there are insufficient, quantitative data on the tanks and waste to prove that. Protection, such as grounding of risers and air terminals on existing light poles, is recommended.

  17. Probability, consequences, and mitigation for lightning strikes to Hanford site high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Zach, J.J.

    1996-08-01

    The purpose of this report is to summarize selected lightning issues concerning the Hanford Waste Tanks. These issues include the probability of lightning discharge striking the area immediately adjacent to a tank including a riser, the consequences of significant energy deposition from a lightning strike in a tank, and mitigating actions that have been or are being taken. The major conclusion of this report is that the probability of a lightning strike depositing sufficient energy in a tank to cause an effect on employees or the public is unlikely;but there are insufficient, quantitative data on the tanks and waste to prove that. Protection, such as grounding of risers and air terminals on existing light poles, is recommended.

  18. High-level waste tank remediation technology integration summary. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    DeLannoy, C.R.; Susiene, C. [Enserch Environmental Inc., Bellevue, WA (United States); Fowler, K.M. [Pacific Northwest Lab., Richland, WA (United States); Robson, W.M. [Lawrence Livermore National Lab., CA (United States); Cruse, J.M. [Westinghouse Hanford Co., Richland, WA (United States)

    1994-07-01

    The U.S. Department of Energy`s Environmental Restoration and Waste Management and Technology Development Programs are engaged in a number of projects to develop, demonstrate, test, and evaluate new technologies to support the cleanup and site remediation of more than 300 underground storage tanks containing over 381,000 m{sup 3} (100 million gal) of liquid radioactive mixed waste at the Hanford Reservation. Significant development is needed within primary functions and in determining an overall bounding strategy. This document is an update of continuing work to summarize the overall strategy and to provide data regarding technology development activities within the strategy. It is intended to serve as an information resource to support understanding, decision making, and integration of multiple program technology development activities. Recipients are encouraged to provide comments and input to the authors for incorporation in future revisions.

  19. Functions and requirements document for interim store solidified high-level and transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Smith-Fewell, M.A., Westinghouse Hanford

    1996-05-17

    The functions, requirements, interfaces, and architectures contained within the Functions and Requirements (F{ampersand}R) Document are based on the information currently contained within the TWRS Functions and Requirements database. The database also documents the set of technically defensible functions and requirements associated with the solidified waste interim storage mission.The F{ampersand}R Document provides a snapshot in time of the technical baseline for the project. The F{ampersand}R document is the product of functional analysis, requirements allocation and architectural structure definition. The technical baseline described in this document is traceable to the TWRS function 4.2.4.1, Interim Store Solidified Waste, and its related requirements, architecture, and interfaces.

  20. Evaluation of alternatives for high-level and transuranic radioactive- waste disposal standards

    Energy Technology Data Exchange (ETDEWEB)

    Klett, R.D. [Sandia National Labs., Albuquerque, NM (United States); Gruebel, M.M. [Tech. Reps., Inc., Albuquerque, NM (United States)

    1992-12-01

    The remand of the US Environmental Protection Agency`s long-term performance standards for radioactive-waste disposal provides an opportunity to suggest modifications that would make the regulation more defensible and remove inconsistencies yet retain the basic structure of the original rule. Proposed modifications are in three specific areas: release and dose limits, probabilistic containment requirements, and transuranic-waste disposal criteria. Examination of the modifications includes discussion of the alternatives, demonstration of methods of development and implementation, comparison of the characteristics, attributes, and deficiencies of possible options within each area, and analysis of the implications for performance assessments. An additional consideration is the impact on the entire regulation when developing or modifying the individual components of the radiological standards.

  1. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  2. Chemical Oxygen Demand (COD) For Monitoring Reduction-Oxidation (Redox) Equilibrium During High Level Waste (HLW) Vitrification

    Energy Technology Data Exchange (ETDEWEB)

    JANTZEN, CAROLM.

    2004-04-30

    High-level nuclear waste is being immobilized at the Savannah River Site by vitrification into borosilicate glass at the Defense Waste Processing Facility. Control of the REDuction/OXidation (REDOX) equilibrium in the DWPF melter is critical for processing high level liquid wastes. Based upon previous research, an acceptable iron REDOX ratio was defined for the DWPF melts as 0.09 Fe2/SFe 0.33. Controlling the DWPF melter at a REDuction/OXidation (REDOX) equilibrium ofFe2/SFe 0.33 prevents the potential for metallic and metallic sulfide species to form and accumulate on the floor of the melter. Control of foaming due to deoxygenation of manganic species is achieved by converting 66-100 of the MnO2 or Mn2O3 species in a waste feed to MnO before the waste is fed to the DWPF melter. At the lower redox limit of Fe 2/SFe 0.09 about 99 of the Mn 4/Mn 3 is converted to Mn 2. Therefore, the lower REDOX limit eliminates melter foaming from deoxygenation. Organic and nitrate concentrations in the DWPF melter feed are the major parameters influencing melt REDOX. Organics such as formates act as reductants while nitrates, nitrites, and manganic (Mn 4 and Mn 3) species act as oxidants. During melting, the REDOX of the melt pool cannot be measured. Therefore, the Fe 2/SFe ratio in the glass poured from the melter must be related to melter feed organic and oxidant concentrations to ensure production of a high quality glass without impacting production rate (e.g., from foaming) or melter life (e.g., from metal formation and accumulation).

  3. Analysis of social recognition and disposal time for the long-term management scenario of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Myong; Jeong, Jin Yeop; Ha, Gyu Man [Korea Energy Technology and Emergency Management Institute, Seoul (Korea, Republic of)

    2009-02-15

    Recently, the need of high-level radioactive waste policy including spent fuel management becomes serious due to the rapid increase in oil price, the nationalism of natural resources, and the environmental issues such as Tokyo protocol. Also, the policy should be established urgently to prepare the saturation of on-site storage capacity of spent fuel, the revision of 'Agreement for Cooperation-Concerning Civil Uses of Atomic Energy' between Korea and US, the anxiety for nuclear weapon proliferation, and R and D to reduce the amount of waste to be disposed. In this study, we performed case study of US, Japan, Canada and Finland, which have special laws and plans/roadmaps for high-level waste management, to draw the policy requirements to be considered in HLW management. Also, we reviewed social conflict issues experienced in our society, and summarized the factors affecting the political and social environment. These policy requirements and factors summarized in this study should be considered seriously in the process for public consensus and the policy making regarding HLW management. Finally, the following 4 action items were drawn to manage HLW successfully : - Continuous and systematic R and D activities to obtain reliable management technology - Promoting companies having specialty in HLW management - Nurturing experts and workforce - Drive the public consensus process

  4. Cement-based radioactive waste hosts formed under elevated temperatures and pressures (FUETAP concretes) for Savannah River Plant high-level defense waste

    Energy Technology Data Exchange (ETDEWEB)

    Dole, L.R.; Rogers, G.C.; Morgan, M.T.; Stinton, D.P.; Kessler, J.H.; Robinson, S.M.; Moore, J.G.

    1983-03-01

    Concretes that are formed under elevated temperatures and pressures (called FUETAP) are effective hosts for high-level radioactive defense wastes. Tailored concretes developed at the Oak Ridge National Laboratory (ORNL) have been prepared from common Portland cements, fly ash, sand, clays, and waste products. These concretes are produced by accelerated curing under mild autoclave conditions (85 to 200/sup 0/C, 0.1 to 1.5 MPa) for 24 h. The solids are subsequently dewatered (to remove unbound water) at 250/sup 0/C for 24 h. The resulting products are strong (compressive strength, 40 to 100 MPa), leach resistant (plutonium leaches at the rate of 10 pg/(cm/sup 2/.d)), and radiolytically stable, monolithic waste forms (total gas value = 0.005 molecule/100 eV). This report summarizes the results of a 4-year FUETAP development program for Savannah River Plant (SRP) high-level defense wastes. It addresses the major questions concerning the performance of concretes as radioactive waste forms. These include leachability, radiation stability, thermal stability, thermal conductivity, impact strength, permeability, phase complexity, and effect of waste composition.

  5. Noble Metals and Spinel Settling in High Level Waste Glass Melters

    Energy Technology Data Exchange (ETDEWEB)

    Sundaram, S. K.; Perez, Joseph M.

    2000-09-30

    In the continuing effort to support the Defense Waste Processing Facility (DWPF), the noble metals issue is addressed. There is an additional concern about the amount of noble metals expected to be present in the future batches that will be considered for vitrification in the DWPF. Several laboratory, as well as melter-scale, studies have been completed by various organizations (mainly PNNL, SRTC, and WVDP in the USA). This letter report statuses the noble metals issue and focuses at the settling of noble metals in melters.

  6. Everything you always wanted to know about shipping high-level nuclear wastes

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    In this document, DOE has gathered together the ''most often asked'' questions, and has furnished a detailed answer to each question. Individual answers are not necessarily totally self-contained, and it may be necessary to put several of the answers together in order to obtain a full picture of the various aspects of a particular question. The answers to the questions then serve a dual purpose--to provide specific information on commonly asked questions, in understandable terms, and to present the composite parts of an overall program on just how the wastes will be transported and the many related factors. Accident situations are covered.

  7. Review of important rock mechanics studies required for underground high level nuclear waste repository program

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S.; Cho, W. J

    2007-01-15

    Disposal concept adapting room and pillar method, which is a confirmed technique in mining and tunnel construction for long time, has advantages at cost, safety, technical feasibility, flexibility, and international cooperation point of views. Then the important rock mechanics principals and in situ and laboratory tests for understanding the behavior of rock, buffer, and backfill as well as their interactions will be reviewed. The accurate understanding of them is important for developing a safe disposal concept and successful operation of underground repository for permanent disposal of radioactive wastes. First of all, In this study, current status of rock mechanics studies for HLW disposal in foreign countries such as Sweden, USA, Canada, Finland, Japan, and France were reviewed. After then the in situ and laboratory tests for site characterization were summarized. Furthermore, rock mechanics studies required during the whole procedure for the disposal project from repository design to the final closure will be reviewed systematically. This study will help for developing a disposal system including site selection, repository design, operation, maintenance, and closure of a repository in deep underground rock. By introducing the required rock mechanics tests at different stages, it would be helpful from the planning stage to the operation stage of a radioactive waste disposal project.

  8. Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    K.G. Mon; F. Hua

    2005-04-12

    This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking, and hydrogen induced cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure.

  9. Development of integraded mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J. C., LLNL

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored.

  10. The high level and long lived radioactive wastes; Les dechets radioactifs a haute activite et a vie longue

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This report presents the main conclusions of 15 years of researches managed by the CEA. This report is the preliminary version of the 2005 final report. It presents the main conclusions of the actions on the axis 1 and 3 of the law of the 30 December 1991. The synthesis report on the axis 1 concerns results obtained on the long lived radionuclides separation and transmutation in high level and long lived radioactive wastes. the synthesis report on the axis 3 presents results obtained by the processes of conditioning and of ground and underground long term storage. (A.L.B.)

  11. Comments on a paper tilted `The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues`

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; McConnell, P.E.; Nigrey, P.J.; Ammerman, D.J. [and others

    1997-05-01

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed.

  12. Preliminary total-system analysis of a potential high-level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Eslinger, P.W.; Doremus, L.A.; Engel, D.W.; Miley, T.B.; Murphy, M.T.; Nichols, W.E.; White, M.D. [Pacific Northwest Lab., Richland, WA (United States); Langford, D.W.; Ouderkirk, S.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-01-01

    The placement of high-level radioactive wastes in mined repositories deep underground is considered a disposal method that would effectively isolate these wastes from the environment for long periods of time. This report describes modeling performed at PNL for Yucca Mountain between May and November 1991 addressing the performance of the entire repository system related to regulatory criteria established by the EPA in 40 CFR Part 191. The geologic stratigraphy and material properties used in this study were chosen in cooperation with performance assessment modelers at Sandia National Laboratories (SNL). Sandia modeled a similar problem using different computer codes and a different modeling philosophy. Pacific Northwest Laboratory performed a few model runs with very complex models, and SNL performed many runs with much simpler (abstracted) models.

  13. LIFE ESTIMATION OF HIGH LEVEL WASTE TANK STEEL FOR F-TANK FARM CLOSURE PERFORMANCE ASSESSMENT - 9310

    Energy Technology Data Exchange (ETDEWEB)

    Subramanian, K; Bruce Wiersma, B; Stephen Harris, S

    2009-01-12

    High level radioactive waste (HLW) is stored in underground carbon steel storage tanks at the Savannah River Site. The underground tanks will be closed by removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations, and severing/sealing external penetrations. The life of the carbon steel materials of construction in support of the performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to grouted conditions. A stochastic approach was followed to estimate the distributions of failures based upon mechanisms of corrosion accounting for variances in each of the independent variables. The methodology and results used for one-type of tank is presented.

  14. Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Heinz J.

    2013-06-24

    The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

  15. PHYSICAL CHARACTERIZATION OF VITREOUS STATE LABORATORY AY102/C106 AND AZ102 HIGH LEVEL WASTE MELTER FEED SIMULANTS (U)

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, E

    2005-03-31

    The objective of this task is to characterize and report specified physical properties and pH of simulant high level waste (HLW) melter feeds (MF) processed through the scaled melters at Vitreous State Laboratories (VSL). The HLW MF simulants characterized are VSL AZ102 straight hydroxide melter feed, VSL AZ102 straight hydroxide rheology adjusted melter feed, VSL AY102/C106 straight hydroxide melter feed, VSL AY102/C106 straight hydroxide rheology adjusted melter feed, and Savannah River National Laboratory (SRNL) AY102/C106 precipitated hydroxide processed sludge blended with glass former chemicals at VSL to make melter feed. The physical properties and pH were characterized using the methods stated in the Waste Treatment Plant (WTP) characterization procedure (Ref. 7).

  16. Selective removal of cesium and strontium using porous frameworks from high level nuclear waste.

    Science.gov (United States)

    Aguila, Briana; Banerjee, Debasis; Nie, Zimin; Shin, Yongsoon; Ma, Shengqian; Thallapally, Praveen K

    2016-05-01

    Efficient and cost-effective removal of radioactive (137)Cs and (90)Sr found in spent fuel is an important step for safe, long-term storage of nuclear waste. Solid-state materials such as resins and titanosilicate zeolites have been assessed for the removal of Cs and Sr from aqueous solutions, but there is room for improvement in terms of capacity and selectivity. Herein, we report the Cs(+) and Sr(2+) exchange potential of an ultra stable MOF, namely, MIL-101-SO3H, as a function of different contact times, concentrations, pH levels, and in the presence of competing ions. Our preliminary results suggest that MOFs with suitable ion exchange groups can be promising alternate materials for cesium and strontium removal.

  17. Development and application of a deflagration pressure analysis code for high level waste processing

    Energy Technology Data Exchange (ETDEWEB)

    Hensel, S.J.; Thomas, J.K.

    1994-06-01

    The Deflagration Pressure Analysis Code (DPAC) was developed primarily to evaluate peak pressures for deflagrations in radioactive waste storage and process facilities at the Savannah River Site (SRS). Deflagrations in these facilities are generally considered to be incredible events, but it was judged prudent to develop modeling capabilities in order to facilitate risk estimates. DPAC is essentially an engineering analysis tool, as opposed to a detailed thermal hydraulics code. It accounts for mass loss via venting, energy dissipation by radiative heat transfer, and gas PdV work. Volume increases due to vessel deformation can also be included using pressure-volume data from a structural analysis of the enclosure. This paper presents an overview of the code, benchmarking, and applications at SRS.

  18. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  19. Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing

    Energy Technology Data Exchange (ETDEWEB)

    Seymour, R.G.

    1995-06-07

    During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H{sub 2}) and ammonia (NH{sub 3}) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H{sub 2} and NH{sub 3} could evolve at appreciable rates and quantities. The explosive nature of H{sub 2} and NH{sub 3} (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed.

  20. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  1. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

    Energy Technology Data Exchange (ETDEWEB)

    None

    1988-06-01

    The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system as developed under the Nuclear Waste Policy Act of 1982. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C; Michael Stone, M

    2007-03-30

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

  3. Characterizing the proposed geologic repository for high-level radioactive waste at Yucca Mountain, Nevada--hydrology and geochemistry

    Science.gov (United States)

    Stuckless, John S.; Levich, Robert A.

    2012-01-01

    This hydrology and geochemistry volume is a companion volume to the 2007 Geological Society of America Memoir 199, The Geology and Climatology of Yucca Mountain and Vicinity, Southern Nevada and California, edited by Stuckless and Levich. The work in both volumes was originally reported in the U.S. Department of Energy regulatory document Yucca Mountain Site Description, for the site characterization study of Yucca Mountain, Nevada, as the proposed U.S. geologic repository for high-level radioactive waste. The selection of Yucca Mountain resulted from a nationwide search and numerous committee studies during a period of more than 40 yr. The waste, largely from commercial nuclear power reactors and the government's nuclear weapons programs, is characterized by intense penetrating radiation and high heat production, and, therefore, it must be isolated from the biosphere for tens of thousands of years. The extensive, unique, and often innovative geoscience investigations conducted at Yucca Mountain for more than 20 yr make it one of the most thoroughly studied geologic features on Earth. The results of these investigations contribute extensive knowledge to the hydrologic and geochemical aspects of radioactive waste disposal in the unsaturated zone. The science, analyses, and interpretations are important not only to Yucca Mountain, but also to the assessment of other sites or alternative processes that may be considered for waste disposal in the future. Groundwater conditions, processes, and geochemistry, especially in combination with the heat from radionuclide decay, are integral to the ability of a repository to isolate waste. Hydrology and geochemistry are discussed here in chapters on unsaturated zone hydrology, saturated zone hydrology, paleohydrology, hydrochemistry, radionuclide transport, and thermally driven coupled processes affecting long-term waste isolation. This introductory chapter reviews some of the reasons for choosing to study Yucca Mountain as a

  4. Characterizing the proposed geologic repository for high-level radioactive waste at Yucca Mountain, Nevada: hydrology and geochemistry

    Science.gov (United States)

    Stuckless, John S.; Levich, Robert A.

    2012-01-01

    This hydrology and geochemistry volume is a companion volume to the 2007 Geological Society of America Memoir 199, The Geology and Climatology of Yucca Mountain and Vicinity, Southern Nevada and California, edited by Stuckless and Levich. The work in both volumes was originally reported in the U.S. Department of Energy regulatory document Yucca Mountain Site Description, for the site characterization study of Yucca Mountain, Nevada, as the proposed U.S. geologic repository for high-level radioactive waste. The selection of Yucca Mountain resulted from a nationwide search and numerous committee studies during a period of more than 40 yr. The waste, largely from commercial nuclear power reactors and the government's nuclear weapons programs, is characterized by intense penetrating radiation and high heat production, and, therefore, it must be isolated from the biosphere for tens of thousands of years. The extensive, unique, and often innovative geoscience investigations conducted at Yucca Mountain for more than 20 yr make it one of the most thoroughly studied geologic features on Earth. The results of these investigations contribute extensive knowledge to the hydrologic and geochemical aspects of radioactive waste disposal in the unsaturated zone. The science, analyses, and interpretations are important not only to Yucca Mountain, but also to the assessment of other sites or alternative processes that may be considered for waste disposal in the future. Groundwater conditions, processes, and geochemistry, especially in combination with the heat from radionuclide decay, are integral to the ability of a repository to isolate waste. Hydrology and geochemistry are discussed here in chapters on unsaturated zone hydrology, saturated zone hydrology, paleohydrology, hydrochemistry, radionuclide transport, and thermally driven coupled processes affecting long-term waste isolation. This introductory chapter reviews some of the reasons for choosing to study Yucca Mountain as a

  5. Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Kovach, L.A. [ed.] [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Murphy, W.M. [ed.] [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1995-09-01

    A Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste was held in San Antonio, Texas on July 22--25, 1991. The proceedings comprise seventeen papers submitted by participants at the workshop. A series of papers addresses the relation of natural analog studies to the regulation, performance assessment, and licensing of a geologic repository. Applications of reasoning by analogy are illustrated in papers on the role of natural analogs in studies of earthquakes, petroleum, and mineral exploration. A summary is provided of a recently completed, internationally coordinated natural analog study at Pocos de Caldas, Brazil. Papers also cover problems and applications of natural analog studies in four technical areas of nuclear waste management-. waste form and waste package, near-field processes and environment, far-field processes and environment, and volcanism and tectonics. Summaries of working group deliberations in these four technical areas provide reviews and proposals for natural analog applications. Individual papers have been cataloged separately.

  6. Experiences from risk communication in the siting of a geological repository for high level waste in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Thegerstroem, C.; Engstroem, S. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    1999-12-01

    SKB is planning in the year 2001 to designate two siting alternatives for further site characterisation. The work in the municipalities of Oesthammar, Nykoeping, Oskarshamn and Tierp is taking place in an atmosphere of constructive discussions. There is a growing feeling in Sweden among broad categories of the public that the nuclear waste exists and should be taken care of by our generation, without many of these people ever getting positive to the use of nuclear energy. While the NIMBY syndrome might still have a good grip on some, there has never been a more constructive debate about the nuclear waste than now, even though there still is a lot of work to do. Siting a geological repository for high level waste puts our democratic system under hard tests. The decision making process is about openness, skills in interacting with the public, respect of people's fears and concerns and at last but not the least independent, competent and visible participation by other stakeholders (politicians locally and nationally, regulatory bodies etc). Good skills in risk communication are important ingredients that might facilitate SKB's task as a developer. Far more important however, is the trust we might get from past and present record of handling the waste and from the way we work and behave in the feasibility studies in the municipalities where SKB is involved.

  7. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  8. Modeling for speciation of radionuclides in waste packages with high-level radioactive wastes; Modellierung zur Speziation von Radionukliden in Abfallgebinden mit hoch radioaktiven Abfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Weyand, Torben; Bracke, Guido; Seher, Holger

    2016-10-15

    Based on a literature search on radioactive waste inventories adequate thermodynamic data for model inventories were derived for geochemical model calculations using PHREEQC in order to determine the solid phase composition of high-level radioactive wastes in different containers. The calculations were performed for different model inventories (PWR-MOX, PWR-UO2, BWR-MOX, BMR-UO2) assuming intact containers under reduction conditions. The effect of a defect in the container on the solid phase composition was considered in variation calculations assuming air contact induced oxidation.

  9. Biohydrogen Fermentation from Sucrose and Piggery Waste with High Levels of Bicarbonate Alkalinity

    Directory of Open Access Journals (Sweden)

    Jeongdong Choi

    2015-03-01

    Full Text Available This study examined the influence of biohydrogen fermentation under the high bicarbonate alkalinity (BA and pH to optimize these critical parameters. When sucrose was used as a substrate, hydrogen was produced over a wide range of pH values (5–9 under no BA supplementation; however, BA affected hydrogen yield significantly under different initial pHs (5–10. The actual effect of high BA using raw piggery waste (pH 8.7 and BA 8.9 g CaCO3/L showed no biogas production or propionate/acetate accumulation. The maximum hydrogen production rate (0.32 L H2/g volatile suspended solids (VSS-d was observed at pH 8.95 and 3.18 g CaCO3/L. BA greater than 4 g CaCO3/L also triggered lactate-type fermentation, leading to propionate accumulation, butyrate reduction and homoacetogenesis, potentially halting the hydrogen production rate. These results highlight that the substrate with high BA need to amend adequately to maximize hydrogen production.

  10. Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, Bruce A.; Bazelaire, Eve; Bonnesen, Peter V.; Custelcean, Radu; Delmau, Laetitia H.; Ditto, Mary E.; Engle, Nancy L.; Gorbunova, Maryna G.; Haverlock, Tamara J.; Levitskaia, Tatiana G.; Bartsch, Richard A.; Surowiec, Malgorzata A.; Zhou, Hui

    2005-07-06

    This project unites expertise at Oak Ridge National Laboratory (ORNL) and Texas Tech University (TTU, Prof. Richard A. Bartsch) to answer fundamental questions addressing the problem of cesium removal from high-level tank waste. Efforts focus on novel solvent-extraction systems containing calixcrown extractants designed for enhanced cesium binding and release. Exciting results are being obtained in three areas: (1) a new lipophilic cesium extractant with a high solubility in the solvent; (2) new proton-ionizable calixcrowns that both strongly extract cesium and ''switch off'' when protonated; and (3) an improved solvent system that may be stripped with more than 100-fold greater efficiency. Scientific questions primarily concern how to more effectively reverse extraction, focusing on the use of amino groups and proton-ionizable groups to enable pH-switching. Synthesis is being performed at ORNL (amino calixcrowns) and TTU (proton-ionizable calixcrowns). At ORNL, the extraction behavior is being surveyed to assess the effectiveness of candidate solvent systems, and systematic distribution measurements are under way to obtain a thermodynamic understanding of partitioning and complexation equilibria. Crystal structures obtained at ORNL are revealing the structural details of cesium binding. The overall objective is a significant advance in the predictability and efficiency of cesium extraction from high-level waste in support of potential implementation at U. S. Department of Energy (USDOE) sites.

  11. Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, Bruce A; Bazelaire, Eve; Bonnesen, Peter V; Custelcean, Radu; Delmau, Laetitia H; Ditto, Mary E; Engle, Nancy L; Gorbunova, Maryna G; Haverlock, Tamara J; Levitskaia, Taiana G; Bartsch, Richard A; Surowiec, Malgorzata A; Zhou, Hui

    2005-07-06

    This project unites expertise at Oak Ridge National Laboratory (ORNL) and Texas Tech University (TTU, Prof. Richard A. Bartsch) to answer fundamental questions addressing the problem of cesium removal from high-level tank waste. Efforts focus on novel solvent-extraction systems containing calixcrown extractants designed for enhanced cesium binding and release. Exciting results are being obtained in three areas: (1) a new lipophilic cesium extractant with a high solubility in the solvent; (2) new proton-ionizable calixcrowns that both strongly extract cesium and "switch off" when protonated; and (3) an improved solvent system that may be stripped with more than 100-fold greater efficiency. Scientific questions primarily concern how to more effectively reverse extraction, focusing on the use of amino groups and proton-ionizable groups to enable pH-switching. Synthesis is being performed at ORNL (amino calixcrowns) and TTU (proton-ionizable calixcrowns). At ORNL, the extraction behavior is being surveyed to assess the effectiveness of candidate solvent systems, and systematic distribution measurements are under way to obtain a thermodynamic understanding of partitioning and complexation equilibria. Crystal structures obtained at ORNL are revealing the structural details of cesium binding. The overall objective is a significant advance in the predictability and efficiency of cesium extraction from high-level waste in support of potential implementation at U. S. Department of Energy (USDOE) sites.

  12. Survey and analysis of the domestic technology level for the concept development of high level waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byung Su; Song, Jae Hyok [Seoul National University, Seoul (Korea); Park, Kwang Hon; Hwang, Ju Ho; Park, Sung Hyun; Lee, Jae Min [Kyunghee University, Seoul (Korea); Han, Joung Sang; Kim, Ku Young [Yonsei University, Seoul (Korea); Lee, Jae Ki; Chang, Jae Kwon [Hangyang University, Seoul (Korea)

    1998-09-01

    The objectives of this study are the analysis of the status of HLW disposal technology and the investigation of the domestic technology level. The study has taken two years to complete with the participation of forty five researchers. The study was mainly carried out through means of literature surveys, collection of related data, visits to research institutes, and meetings with experts in the specific fields. During the first year of this project, the International Symposium on the Concept Development of the High Level Waste Disposal System was held in Taejon, Korea in October, 1997. Eight highly professed foreign experts whose fields of expertise projected to the area of high level waste disposal were invited to the symposium. This study is composed of four major areas; disposal system design/construction, engineered barrier characterization, geologic environment evaluation and performance assessment and total safety. A technical tree scheme of HLW disposal has been illustrated according to the investigation and an analysis for each technical area. For each detailed technology, research projects, performing organization/method and techniques that are to be secured in the order of priority are proposed, but the suggestions are merely at a superfluous level of propositional idea due to the reduction of the budget in the second year. The detailed programs on HLW disposal are greatly affected by governmental HLW disposal policy and in this study, the primary decisions to be made in each level of HLW disposal enterprise and a rough scheme are proposed. (author). 20 refs., 97 figs., 33 tabs.

  13. Risk-informed assessment of radionuclide release from dissolution of spent nuclear fuel and high-level waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Tae M., E-mail: tae.ahn@nrc.gov

    2017-06-15

    Highlights: • Dissolution of HLW waste form was assessed with long-term risk informed approach. • The radionuclide release rate decreases with time from the initial release rate. • Fast release radionuclides can be dispersed with discrete container failure time. • Fast release radionuclides can be restricted by container opening area. • Dissolved radionuclides may be further sequestered by sorption or others means. - Abstract: This paper aims to detail the different parameters to be considered for use in an assessment of radionuclide release. The dissolution of spent nuclear fuel and high-level nuclear waste glass was considered for risk and performance insights in a generic disposal system for more than 100,000 years. The probabilistic performance assessment includes the waste form, container, geology, and hydrology. Based on the author’s previous extended work and data from the literature, this paper presents more detailed specific cases of (1) the time dependence of radionuclide release, (2) radionuclide release coupled with container failure (rate-limiting process), (3) radionuclide release through the opening area of the container and cladding, and (4) sequestration of radionuclides in the near field after container failure. These cases are better understood for risk and performance insights. The dissolved amount of waste form is not linear with time but is higher at first. The radionuclide release rate from waste form dissolution can be constrained by container failure time. The partial opening area of the container surface may decrease radionuclide release. Radionuclides sequestered by various chemical reactions in the near field of a failed container may become stable with time as the radiation level decreases with time.

  14. Initial demonstration of the NRC`s capability to conduct a performance assessment for a High-Level Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Codell, R.; Eisenberg, N.; Fehringer, D.; Ford, W.; Margulies, T.; McCartin, T.; Park, J.; Randall, J.

    1992-05-01

    In order to better review licensing submittals for a High-Level Waste Repository, the US Nuclear Regulatory Commission staff has expanded and improved its capability to conduct performance assessments. This report documents an initial demonstration of this capability. The demonstration made use of the limited data from Yucca Mountain, Nevada to investigate a small set of scenario classes. Models of release and transport of radionuclides from a repository via the groundwater and direct release pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year simulation time. Latin hypercube sampling of input parameters was used to express results as distributions and to investigate model sensitivities. This methodology demonstration should not be interpreted as an estimate of performance of the proposed repository at Yucca Mountain, Nevada. By expanding and developing the NRC staff capability to conduct such analyses, NRC would be better able to conduct an independent technical review of the US Department of Energy (DOE) licensing submittals for a high-level waste (HLW) repository. These activities were divided initially into Phase 1 and Phase 2 activities. Additional phases may follow as part of a program of iterative performance assessment at the NRC. The NRC staff conducted Phase 1 activities primarily in CY 1989 with minimal participation from NRC contractors. The Phase 2 activities were to involve NRC contractors actively and to provide for the transfer of technology. The Phase 2 activities are scheduled to start in CY 1990, to allow Sandia National Laboratories to complete development and transfer of computer codes and the Center for Nuclear Waste Regulatory Analyses (CNWRA) to be in a position to assist in the acquisition of the codes.

  15. Closure development for high-level nuclear waste containers for the tuff repository; Phase 1, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Robitz, E.S. Jr.; McAninch, M.D. Jr.; Edmonds, D.P. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01

    This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs.

  16. The Role of Temperature in the Safety Case for High-Level Radioactive Waste Disposal: A Comparison of Design Concepts

    Directory of Open Access Journals (Sweden)

    Joachim Heierli

    2017-06-01

    Full Text Available The disposal of heat-generating radioactive waste in deep underground facilities requires a sparing use of spatial resources on the one side and favorable temperature conditions over the project lifetime on the other side. Under heat-sensitive conditions, these goals run in opposite directions and therefore a balance of some kind must be found. Often the elected strategy is to determine the size of the repository by capping the temperatures in the near-field, thus setting an upper limit to the deterioration of barrier materials. Alternatively, the spatial resources available in the siting area can be used to further reduce temperatures as long as supplementary benefits are returned from doing so. Using analytical modeling of the heat flow in the circumambient rock of a repository for high-level waste and spent fuel, this contribution examines possible obstacles in substantiating the safety case, namely the retrievability of waste during the operational lifetime of the facility, the representativeness of pilot disposal areas for monitoring, and the effect of thermal anomalies underground. The results indicate that there are, amongst the visited criteria, several benefits to the temperature-optimizing strategy over the prevailing space-optimizing concepts. The right balance between saving spatial resources and obtaining optimal temperature conditions is yet to be found.

  17. Main organic materials in a repository for high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta [Vita vegrandis, Hindaas (Sweden); Grive, Mireia; Gaona, Xavier; Duro, Lara; Bruno, Jordi [Enviros Consulting, Valldoreix, Barcelona (Spain)

    2007-11-15

    , although the presence of aromatic compounds and PAHs in groundwater is not desirable by itself, they are of no consequence for the long-term performance of the repository. 5. Detergents and lubricants. The same reasoning as for fuels and engine emissions can be applied in this case. The amount of detergents should be minimized, although in the amounts that they are expected to occur, no important impact is foreseen. 6. Materials from human activities. Among them, the ones having potentially a more important effect are fibres from clothes, due to the presence of cellulose, and therefore it is recommended to minimise human-related wastes, although no large amounts of these materials are expected to be present after the repository closure. The effects that organic substances can have in the repository will always depend on the amounts present in the repository after closure. The estimated average concentrations are below 1.7x10{sup -4} kg/m{sup 3} (0.17 mg/L) of hydrocarbons in the deposition tunnels and less than 8.4x10{sup -4} kg/m{sup 3} (0.84 mg/L) of carbohydrates, assuming a total saturation in the pore water and an even distribution of the organic materials. This should be compared to the organic material found in groundwater at natural circumstances. At 500 m depth the DOC (dissolved organic carbon) content usually are approximately 0.5.2 mg/L. Three processes are deemed to have the largest possible impact on the performance of the repository: i) Increase of the reducing capacity and decrease of the redox potential in the short-term, and increased rate of depletion of the oxygen trapped during the repository operation stage. ii) Increase in the complexing capacity of the groundwater due to the presence of organic complexants, which is expected to be a process of more relevance in the long-term. Many organic molecules with complexing capacity, such as short organic acids like acetate are, however, oxidised as a consequence of microbial metabolism. The acetate

  18. Yucca Mountain or: How we learn to stop worrying and love the Department of Energy`s high-level waste disposal guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, N.

    1993-12-31

    Having inherited the transgressions of our forebears, we face a growing problem: the safe disposal of high-level waste, an intensely radioactive byproduct of the production of nuclear weapons and the operation of nuclear power generators. Roughly 20,000 metric tons of spent fuel (a subset of high-level waste, as the term is used in this Note) are currently stored in canisters in cooling pools beside American power plants. Authorities claim that this amount will double by the year 2000, triple by 2010 and quadruple by 2020. Current storage practice is now recognized as a go-between; the necessarily prolific nature of past and future high-level waste production (both public and private) demands that a permanent disposal solution be found. High-level waste remains hazardously toxic for at least 10,000 years.

  19. Seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1993-01-01

    This document provides guidelines for the design and evaluation of underground high-level waste storage tanks due to seismic loads. Attempts were made to reflect the knowledge acquired in the last two decades in the areas of defining the ground motion and calculating hydrodynamic loads and dynamic soil pressures for underground tank structures. The application of the analysis approach is illustrated with an example. The guidelines are developed for specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document.

  20. A preliminary study on the regional fracture systems for deep geological disposal of high level radioactive waste in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Koh, Young Kown; Park, Byoung Yoon [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    For the deep geological disposal of high-level radioactive waste, it is essential to characterize the fracture system in rock mass which has a potential pathways of nuclide. Currently, none of research results are in classification and detailed properties for the fracture system in Korea. This study aims to classify and describe the regional fracture system in lithological and geotectonical point of view using literature review, shaded relief map, and aeromagnetic survey data. This report contains the following: - Theoretical review of the fracture development mechanism. - Overall fault and fracture map. - Geological description on the distributional characteristics of faults and fractures(zone) in terms of lithological domain and tectonical province. 122 refs., 22 figs., 4 tabs. (Author)

  1. AN ANALYSIS OF THE THERMAL AND MECHANICAL BEHAVIOR OF ENGINEERED BARRIERS IN A HIGH-LEVEL RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    S. KWON

    2013-02-01

    Full Text Available Adequate design of engineered barriers, including canister, buffer and backfill, is important for the safe disposal of high-level radioactive waste. Three-dimensional computer simulations were carried out under different condition to examine the thermal and mechanical behavior of engineered barriers and rock mass. The research looked at five areas of importance, the effect of the swelling pressure, water content of buffer, density of compacted bentonite, emplacement type and the selection of failure criteria. The results highlighted the need to consider tensile stress in the outer shell of a canister due to thermal expansion of the canister and the swelling pressure from the buffer for a more reliable design of an underground repository system. In addition, an adequate failure criterion should be used for the buffer and backfill.

  2. A preliminary study on the geochemical environment for deep geological disposal of high level radioactive waste in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Koh, Yong Kwon; Park, Byoung Yun

    2000-03-01

    Geochemical study on the groundwater from crystalline rocks (granite and gneiss) for the deep geological disposal of high-level radioactive waste was carried out in order to elucidate the hydrogeochemical and isotope characteristics and geochemical evolution of the groundwater. Study areas are Jungwon, Chojeong, Youngcheon and Yusung for granite region, Cheongyang for gneiss region, and Yeosu for volcanic region. Groundwaters of each study areas weree sampled and analysed systematically. Groundwaters can be grouped by their chemistry and host rock. Origin of the groundwater was proposed by isotope ({sup 18}O, {sup 2}H, {sup 13}C, {sup 34}S, {sup 87}Sr, {sup 15}N) studies and the age of groundwater was inferred from their tritium contents. Based ont the geochemical and isotope characteristics, the geochemical evolutions of each types of groundwater were simulated using SOLVEQ/CHILLER and PHREEQC programs.

  3. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms.

  4. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    Energy Technology Data Exchange (ETDEWEB)

    Seymour, R.G.

    1995-02-17

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing.

  5. A natural analogue for high-level waste in tuff: Chemical analysis and modeling of the Valles site

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, H.W.; Krumhansl, J.L.; Ho, C.K. [Sandia National Labs., Albuquerque, NM (United States); Kovach, L. [US Nuclear Regulatory Commission, Washington, DC (United States); McConnell, V.S. [Univ. of Alaska, Fairbanks, AK (United States)

    1995-03-01

    The contact between an obsidian flow and a steep-walled tuff canyon was examined as an analogue for a high-level waste repository. The analogue site is located in the Valles Caldera in New Mexico, where a massive obsidian flow filled a paleocanyon in the Battleship Rock Tuff. The obsidian flow provided a heat source, analogous to waste panels or an igneous intrusion in a repository, and caused evaporation and migration of water. The tuff and obsidian samples were analyzed for major and trace elements and mineralogy by INAA, XRF, x-ray diffraction, and scanning electron microscopy and electron microprobe. Samples were also analyzed for D/H and {sup 39}Ar/{sup 40}Ar isotopic composition. Overall, the effects of the heating event seem to have been slight and limited to the tuff nearest the contact. There is some evidence of devitrification and migration of volatiles in the tuff within 10 m of the contact, but variations in major and trace element chemistry are small and difficult to distinguish from the natural (pre-heating) variability of the rocks.

  6. Annual report, spring 2015. Alternative chemical cleaning methods for high level waste tanks-corrosion test results

    Energy Technology Data Exchange (ETDEWEB)

    Wyrwas, R. B. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-06

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludge in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.

  7. Control of high-level radioactive waste-glass melters. Part 4, Preliminary analysis of DWPF process laboratory capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Bickford, D.F.; Coleman, C.J.

    1990-12-31

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter.

  8. Waste disposal options report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

    1998-02-01

    Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-13

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C; James Marra, J

    2007-09-17

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

  11. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  12. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  13. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  14. Site-specific evaluation of safety issues for high-level waste disposal in crystalline rocks. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jobmann, M. (ed.) [DBE Technology GmbH, Peine (Germany)

    2016-03-31

    In the past, German research and development (R and D) activities regarding the disposal of radioactive waste, including spent nuclear fuel, focused mainly on domal rock salt because rock salt was the preferred host rock formation. In addition, generic R and D work regarding alternative host rocks (crystalline rocks and claystones) had been performed as well for a long time but with lower intensity. Around the year 2000, as a consequence of the moratorium on the Gorleben site, the Federal Government decided to have argillaceous rocks and crystalline rocks investigated in more detail. As Germany does not have any underground research and host rock characterization facilities, international cooperation received a high priority in the German R and D programme for high-level waste (HLW) disposal in order to increase the knowledge regarding alternative host rocks. Major cornerstones of the cooperation are joint projects and experiments conducted especially in underground research laboratories (URL) in crystalline rocks at the Grimsel Test Site (Switzerland) and the Hard Rock Laboratory (HRL) Aespoe(Sweden) and in argillaceous rocks at the URL Mont Terri (Switzerland) and Bure (France). In 2001, the topic of radioactive waste disposal was integrated into the agreement between the former Russian Ministry of Atomic Energy (Minatom, now Rosatom) and the German Ministry of Labor (BMWA), now Ministry of Economic Affairs and Energy (BMWi), on cooperation regarding R and D on the peaceful utilization of nuclear power (agreement on ''Wirtschaftlich-Technische Zusammenarbeit'' WTZ). The intention was to have a new and interesting opportunity for international R and D cooperation regarding HLW disposal in crystalline rocks and the unique possibility to perform site-specific work, to test the safety demonstration tools available, and to expand the knowledge to all aspects specific to these host rocks. Another motivation for joining this cooperation was the

  15. Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues.

    Energy Technology Data Exchange (ETDEWEB)

    Banano, E.J. [Beta Corporation International, Albuquerque, NM (United States); Baca, R.G. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1995-08-01

    The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ``scenario approach,`` while common is not the only method to deal with this uncertainty; other method ``ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ``scenario`` and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository.

  16. Adsorption of Ruthenium, Rhodium and Palladium from Simulated High-Level Liquid Waste by Highly Functional Xerogel - 13286

    Energy Technology Data Exchange (ETDEWEB)

    Onishi, Takashi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Koyama, Shin-ichi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Mimura, Hitoshi [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)

    2013-07-01

    Fission products are generated by fission reactions in nuclear fuel. Platinum group (Pt-G) elements, such as palladium (Pd), rhodium (Rh) and ruthenium (Ru), are also produced. Generally, Pt-G elements play important roles in chemical and electrical industries. Highly functional xerogels have been developed for recovery of these useful Pt-G elements from high - level radioactive liquid waste (HLLW). An adsorption experiment from simulated HLLW was done by the column method to study the selective adsorption of Pt-G elements, and it was found that not only Pd, Rh and Ru, but also nickel, zirconium and tellurium were adsorbed. All other elements were not adsorbed. Adsorbed Pd was recovered by washing the xerogel-packed column with thiourea solution and thiourea - nitric acid mixed solution in an elution experiment. Thiourea can be a poison for automotive exhaust emission system catalysts, so it is necessary to consider its removal. Thermal decomposition and an acid digestion treatment were conducted to remove sulfur in the recovered Pd fraction. The relative content of sulfur to Pd was decreased from 858 to 0.02 after the treatment. These results will contribute to design of the Pt-G element separation system. (authors)

  17. National survey of crystalline rocks and recommendations of regions to be explored for high-level radioactive waste repository sites

    Energy Technology Data Exchange (ETDEWEB)

    Smedes, H.W.

    1983-04-01

    A reconnaissance of the geological literature on large regions of exposed crystalline rocks in the United States provides the basis for evaluating if any of those regions warrant further exploration toward identifying potential sites for development of a high-level radioactive waste repository. The reconnaissance does not serve as a detailed evaluation of regions or of any smaller subunits within the regions. Site performance criteria were selected and applied insofar as a national data base exists, and guidelines were adopted that relate the data to those criteria. The criteria include consideration of size, vertical movements, faulting, earthquakes, seismically induced ground motion, Quaternary volcanic rocks, mineral deposits, high-temperature convective ground-water systems, hydraulic gradients, and erosion. Brief summaries of each major region of exposed crystalline rock, and national maps of relevant data provided the means for applying the guidelines and for recommending regions for further study. It is concluded that there is a reasonable likelihood that geologically suitable repository sites exist in each of the major regions of crystalline rocks. The recommendation is made that further studies first be conducted of the Lake Superior, Northern Appalachian and Adirondack, and the Southern Appalachian Regions. It is believed that those regions could be explored more effectively and suitable sites probably could be found, characterized, verified, and licensed more readily there than in the other regions.

  18. Evaluation of long-term behavior of concretes in high level waste repositories. An accelerated leaching test

    Directory of Open Access Journals (Sweden)

    Hidalgo, A.

    2004-04-01

    Full Text Available The present work describes an accelerated leaching method that with a rapid process allows to develop and evaluate cements for use in a nuclear disposal, and the understanding of the long term effects. The method has been developed to study the stability of cementitious materials in contact with bentonite, to be used in high level radioactivity waste repositories. Nitric acid has been selected to simulate in an accelerated way the pH decreasing produced when concrete is in contact with groundwaters.

    El presente trabajo describe un ensayo acelerado de lixiviación, que mediante un proceso rápido, permite desarrollar y evaluar cementos para su uso en instalaciones nucleares, y la comprensión de su comportamiento a largo plazo. El método se ha desarrollado para estudiar la estabilidad de materiales de base cemento, en contacto con bentonita, que serán utilizados en almacenamientos de resíduos radiactivos de alta actividad. Como agente lixiviante se seleccionó el ácido nítrico, con objeto de simular de forma acelerada, la disminución del pH que se produce cuando el hormigón entra en contacto con aguas subterráneas.

  19. Prediction of corrosion depth of selected materials for the container of high-level wastes under a repository condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung; Choi, Jong Won; Han, Kyung Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    The corrosion depth of selected materials for container of high-level wastes in an underground disposal condition was predicted by analyzing the corrosion behaviors and corrosion rates of copper/copper alloys, carbon steel, titanium/titanium alloys, stainless steel and nickel alloys. Their corrosion rates depend on the amount of oxygen and microbes in bentonite at the bore hole, and local corrosion in addition to general corrosion. However, the effect of radiation and the oxygen dissolved in groundwater would be insignificant. To calculate the corrosion depth, it is assumed that the total amount of oxygen contained in the pore and surface of a bentonite block, and in the gaps among container, rock and bentonite block at a borehole is 300 moles. Assuming that all organic compounds in a bentonite block are presumed as lactate, they would produce 2,100 moles of HS-. The corrosion depths were calculated based on the above assumptions and the wall thickness of copper, carbon steel, titanium, stainless steel and nickel alloys of at least 2.6, 25, 1.3, 5 and 0.3 mm would be required for their corrosion allowances that guarantee their desired service life of 1,000 years. 94 refs., 13 figs., 6 tabs. (Author)

  20. FTIR and Mössbauer spectroscopic study of sodium–aluminum–iron phosphate glassy materials for high level waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, S.V., E-mail: serge.stefanovsky@yandex.ru [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Stefanovsky, O.I. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Remizov, M.B.; Belanova, E.A.; Kozlov, P.V. [FSUE PA Mayak, Central Plant Laboratory, Ozersk, Chelyabinsk Reg. (Russian Federation); Glazkova, Ya.S.; Sobolev, A.V.; Presniakov, I.A. [Lomonosov Moscow State University, Department of Radiochemistry (Russian Federation); Kalmykov, S.N. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Lomonosov Moscow State University, Department of Radiochemistry (Russian Federation); Vernadsky Institute of Geochemistry and Analytical Chemistry of the Russian Academy of Sciences, Laboratory of Radiochemistry, Moscow (Russian Federation); Myasoedov, B.F. [Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences, Laboratory of Radioecology and Radiation Problems, Moscow (Russian Federation); Vernadsky Institute of Geochemistry and Analytical Chemistry of the Russian Academy of Sciences, Laboratory of Radiochemistry, Moscow (Russian Federation)

    2015-11-15

    Complex sodium-aluminum-iron phosphate glassy materials with various Al{sub 2}O{sub 3} to Fe{sub 2}O{sub 3} ratio containing high level waste (HLW) surrogate were characterized by X-ray diffraction and scanning electron microscopy and studied in details by Fourier transform infrared (FTIR) spectroscopy. The samples with high Al{sub 2}O{sub 3} content and not containing Fe{sub 2}O{sub 3} were predominantly amorphous but subjected to devitrification under annealing. Addition of B{sub 2}O{sub 3} and partial Fe{sub 2}O{sub 3} substitution for Al{sub 2}O{sub 3} in the materials increases their resistance to devitrification whereas further substitution and NiO incorporation significantly increase the tendency to devitrification. FTIR spectra demonstrate changes in the structure of glassy materials caused by both structural variations in the anionic motif and occurrence of crystalline phases in the materials. According to Mössbauer spectroscopy data, iron in the glassy samples is present as octahedrally coordinated Fe{sup 3+} ions while in the partly devitrified samples iron is partitioned among vitreous and crystalline phases entering the vitreous phase mainly as Fe{sup 3+}O{sub 6} units and crystalline phases as major Fe{sup 3+} and minor Fe{sup 2+} ions in a magnetically ordered state and participating in a “fast” electronic exchange.

  1. Seismic design and evaluation guidelines for the Department of Energy High-Level Waste Storage Tanks and Appurtenances

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1995-10-01

    This document provides seismic design and evaluation guidelines for underground high-level waste storage tanks. The guidelines reflect the knowledge acquired in the last two decades in defining seismic ground motion and calculating hydrodynamic loads, dynamic soil pressures and other loads for underground tank structures, piping and equipment. The application of the guidelines is illustrated with examples. The guidelines are developed for a specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document. The original version of this document was published in January 1993. Since then, additional studies have been performed in several areas and the results are included in this revision. Comments received from the users are also addressed. Fundamental concepts supporting the basic seismic criteria contained in the original version have since then been incorporated and published in DOE-STD-1020-94 and its technical basis documents. This information has been deleted in the current revision.

  2. Modeling pitting corrosion damage of high-level radioactive-waste containers, with emphasis on the stochastic approach

    Energy Technology Data Exchange (ETDEWEB)

    Henshall, G.A.; Halsey, W.G.; Clarke, W.L.; McCright, R.D.

    1993-01-01

    Recent efforts to identify methods of modeling pitting corrosion damage of high-level radioactive-waste containers are described. The need to develop models that can provide information useful to higher level system performance assessment models is emphasized, and examples of how this could be accomplished are described. Work to date has focused upon physically-based phenomenological stochastic models of pit initiation and growth. These models may provide a way to distill information from mechanistic theories in a way that provides the necessary information to the less detailed performance assessment models. Monte Carlo implementations of the stochastic theory have resulted in simulations that are, at least qualitatively, consistent with a wide variety of experimental data. The effects of environment on pitting corrosion have been included in the model using a set of simple phenomenological equations relating the parameters of the stochastic model to key environmental variables. The results suggest that stochastic models might be useful for extrapolating accelerated test data and for predicting the effects of changes in the environment on pit initiation and growth. Preliminary ideas for integrating pitting models with performance assessment models are discussed. These ideas include improving the concept of container ``failure``, and the use of ``rules-of-thumb`` to take information from the detailed process models and provide it to the higher level system and subsystem models. Finally, directions for future work are described, with emphasis on additional experimental work since it is an integral part of the modeling process.

  3. Optical and Microcantilever-Based Sensors for Real-Time In Situ Characterization of High-Level Waste (81924)

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Gilbert M.; Bonnesen, Peter V.; Dabestani, Reza; Thundat, Thomas G.; Walt, David R.; Bryan, Samuel A.

    2004-06-01

    Fundamental research is being conducted to develop sensors for cesium, strontium, and pertechnetate that can be used to analyze these species in high-level waste (HLW) process streams. Sensors for these species will be combined with sensor elements for OH-, K+, and Na+ in an array that will allow interferences to be corrected. Two fundamentally different approaches are being pursued, having in common the dependence on highly selective molecular recognition agents. In one approach, an array of chemically selective sensors with sensitive fluorescent probes to signal the presence of the constituent of interest will be coupled to fiber optics for remote analytical applications. The second approach employs sensitive microcantilever sensors that have been demonstrated to have unprecedented sensitivity in solution. Selectivity in microcantilever-based sensors is achieved by modifying the surface of a gold-coated cantilever with a monolayer coating of an alkanethiol derivative of the molecular recognition agent. The microcantilever-based sensors function by converting molecular complexation into surface stress. The fundamental technology for fiber optic and cantilever sensors has been developed by our collaborators David Walt and Thomas Thundat, respectively, and the goal of this project is to adapt molecular recognition chemistry to the methods already being employed. Molecular recognition with these sensors is achieved using ionophores constructed with the three dimensional architecture provided by calix[4]arenes, a widely used platform for metal ion complexation.

  4. Corrosion of iron steel under sulphate reducing conditions: application to deep geological disposal of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    El Hajj, H.; Abdelouas, A.; Grambow, B. [Nantes Univ., SUBATECH, UMR 6457 IN2P3/CNRS-Ecole des Mines de Nantes, 44 - Nantes (France)

    2009-07-01

    Full text of publication follows: We investigated the corrosion of the steel P235 under disposal conditions of high-level radioactive waste in argillite. The objective was to study the role of sulphate-reducing bacteria (SRB) indigenous to argillite that may develop under excess of hydrogen produced by steel corrosion in groundwater flowing through argillite. Batch and column experiments were conducted in serum bottles/diffusion cells with coupons of P235 steel, argillite and synthetic argillite water. This system provided enough nutrients to grow SRB's within two weeks with corrosion of the steel and formation of iron sulphides as corrosion products. The results of the weight loss method showed that the steel corrosion seems to be faster in the presence of SRB's in comparison to sterilized runs. The mineralogical identification of iron sulphides, observed and analysed with the scanning electron microscope, will be conducted with Raman spectroscopy and X-ray diffraction. The corrosion mechanism will be determined by studying steel cross sections and surface observations after removal of corrosion layer. Experiments with diffusion cells are under way and the collected outlet solutions show a decrease of sulphate concentration with production of sulphide ions. The results to be obtained with the steel corroded in the cell will be presented at the end of the run planned for few months.

  5. Studying the properties and behaviour of high-level radioactive wastes from the BOR-60 reactor U-Pu and U spent fuel experimental gas-fluoride reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kirillovich, A.P.; Lavrinovich, Yu.G.; Vorobej, M.P.; Pimonov, Yu.I.

    1982-07-01

    The results of investigations of physical-chemical and radiation properties of fluoride radioactive wastes produced during experimental reprocessing of spent oxide uranium-plutonium fuel as well as high-level radioactive waste behaviour in the process of their six-year controlled storage are presented. Radioactive gas release from solid wastes, gaseous phase composition, radionuclide leaching rate are determined. Investigations are performed at a special bench. Energy release of the spent fuel and high-level radioactive wastes is determined by means of heat-conducting type calorimeter. Gas mixture composition in containers with wastes is determined by mass-spectrometric method at equilibrium temperature of high-level radioactive product self-heating and at external container heating up to 700 deg C. Thermal-physical characteristics of solid fluoride wastes are found by the differential thermography method under quasistationary heating. The results obtained show that about a half (44.8-60.9%) of fission product radioactivity is concentrated in fluorination wastes, specific heat release of which constitutes 50-52 W/kg, while ..beta..-activity exceeds 550 TBq/kg. Main contribution into ..beta..-activity is made by Ce, /sup 144/Pr, Ru, /sup 106/Rh, Zr, /sup 95/Nb, /sup 137/Cs. With waste storage time increase their thermal stability increases. It is concluded that the investigation results can be used for calculating the conditions of safe storage of high-level radiactive solid fluoride wastes and optimization of the technological process of spent fuel gas-fluoride reprossing.

  6. Geologic and hydrologic considerations for various concepts of high-level radioactive waste disposal in conterminous United States

    Science.gov (United States)

    Ekren, E.B.; Dinwiddie, G.A.; Mytton, J.W.; Thordarson, William; Weir, J.E.; Hinrichs, E.N.; Schroder, L.J.

    1974-01-01

    The purpose of this investigation is to evaluate and identify which geohydrologic environments in conterminous United States are best suited for various concepts or methods of underground disposal of high-level radioactive wastes and to establish geologic and hydrologic criteria that are pertinent to high-level waste disposal. The unproven methods of disposal include (1) a very deep drill hole (30,000-50,000 ft or 9,140-15,240 m), (2) a matrix of (an array of multiple) drill holes (1,000-20,000 ft or 305-6,100 m), (3) a mined chamber (1,000-10,000 ft or 305-3,050 m), (4) a cavity with separate manmade structures (1,000-10,000 ft or 305-3,050 m), and (5) an exploded cavity (2,000-20,000 ft or 610-6,100 m) o The geohydrologic investigation is made on the presumption that the concepts or methods of disposal are technically feasible. Field and laboratory experiments in the future may demonstrate whether or not any of the methods are practical and safe. All the conclusions drawn are tentative pending experimental confirmation. The investigation focuses principally on the geohydrologic possibilities of several methods of disposal in rocks other than salt. Disposal in mined chambers in salt is currently under field investigation, and this disposal method has been intensely investigated and evaluated by various workers under the sponsorship of the Atomic Energy Commission. Of the various geohydrologic factors that must be considered in the selection of optimum waste-disposal sites, the most important is hydrologic isolation to assure that the wastes will be safely contained within a small radius of the emplacement zone. To achieve this degree of hydrologic isolation, the host rock for the wastes must have very low permeability and the site must be virtually free of faults. In addition, the locality should be in (1) an area of low seismic risk where the possibility of large earthquakes rupturing the emplacement zone is very low, (2) where the possibility- of flooding by

  7. Design concepts of definitive disposal for high level radioactive wastes; Conceptos de diseno de disposicion definitiva para desechos radioactivos de alto nivel

    Energy Technology Data Exchange (ETDEWEB)

    Badillo A, V.E.; Alonso V, G. [ININ, Carr. Mexico-Toluca S/N (Km. 36.5) La Marquesa, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: vbadillo@nuclear.inin.mx

    2007-07-01

    It is excessively known the importance about finding a solution for the handling and disposition of radioactive waste of all level. However, the polemic is centered in the administration of high level radioactive waste and the worn out fuel, forgetting that the more important volumes of waste its are generated in the categories of low level wastes or of very low level. Depending on the waste that will be confined and of the costs, several technological modalities of definitive disposition exist, in function of the depth of the confinement. The concept of deep geologic storage, technological option proposed more than 40 years ago, it is a concept of isolation of waste of long half life placed in a deep underground installation dug in geologic formations that are characterized by their high stability and their low flow of underground water. In the last decades, they have registered countless progresses in technical and scientific aspects of the geologic storage, making it a reliable technical solution supported with many years of scientific work carried out by numerous institutions in the entire world. In this work the design concepts that apply some countries for the high level waste disposal that its liberate heat are revised and the different geologic formations that have been considered for the storage of this type of wastes. (Author)

  8. Optical and Microcantilever-Based Sensors for Real-Time In Situ Characterization of High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Braun, Gilbert M.; Bryan, Samuel

    2002-06-01

    Fundamental research is being conducted to develop sensors for strontium that can be used in real-time to characterize high-level waste (HLW) process streams. Two fundamentally different approaches are being pursued, which have in common the dependence on highly selective molecular recognition agents. In one approach, an array of chemically selective sensors with sensitive fluorescent probes to signal the presence of the constituent of interest are coupled to fiber optics for remote analytical applications. The second approach employs sensitive microcantilever sensors that have been demonstrated to have unprecedented sensitivity in solution for Cs+ and CrO4 -. Selectivity in microcantilever-based sensors is achieved by modifying the surface of a gold-coated cantilever with a monolayer coating of an alkanethiol derivative of the molecular recognition agent. The approaches are complementary since fiber optic sensors can be deployed in the highly alkaline environment of HLW, bu t a method of immobilizing a fluorescent molecular recognition agents in a polymer film or bead on the surface of the optical fiber has yet to be demonstrated. The microcantilever-based sensors function by converting molecular complexation into surface stress, and they have been demonstrated to have the requisite sensitivity. However, we will investigate method of protecting Si or SiN microcantilever sensors in the highly alkaline environment of HLW while maintaining high selectivity. One objective of this project is to develop Sr(II) molecular recognition agents with rapidly established equilibria needed for real-time analysis, and initial research will focus on calixarene-crown ethers as a platform. Sensors for alkali metal ions, hydroxide, and temperature will be part of the array of sensor elements that will be demonstrated in this program for both the cantilever and fiber optic sensor approaches.

  9. Optical and Microcantilever-Based Sensors for Real-Time In Situ Characterization of High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Gilbert M.; Bryan, Samuel

    2003-06-01

    Fundamental research is being conducted to develop sensors for cesium and strontium that can be used in real-time to characterize high-level waste (HLW) process streams. Two fundamentally different approaches are being pursued, having in common the dependence on highly selective molecular recognition agents. In one approach, an array of chemically selective sensors with sensitive fluorescent probes to signal the presence of the constituent of interest will be coupled to fiber optics for remote analytical applications. The second approach employs sensitive microcantilever sensors that have been demonstrated to have unprecedented sensitivity in solution for Cs+ and CrO4 -. Selectivity in microcantilever-based sensors is achieved by modifying the surface of a gold-coated cantilever with a monolayer coating of an alkanethiol derivative of the molecular recognition agent. The approaches are complementary since fiber optic sensors can be deployed in the highly alkaline environment of HLW, but a method of immobilizing a fluorescent molecular recognition agent in a polymer film or bead on the surface of the optical fiber has yet to be demonstrated. The microcantilever-based sensors function by converting molecular complexation into surface stress, and they have been demonstrated to have the requisite sensitivity. However, a method of protecting Si or SiN microcantilever sensors in the highly alkaline environment of HLW while maintaining high selectivity remains to be demonstrated. The fundamental technology for fiber optic and cantilever sensors has been developed by our collaborators David Walt and Thomas Thundat, respectively, and the goal of this project is to adapt molecular recognition chemistry to the methods already being employed. To develop molecular recognition agents for Cs+ and Sr(II) with rapidly established equilibria needed for real-time analysis, we will focus on calixarene-crown ethers as a platform. Sensors for alkali metal ions, hydroxide, and

  10. Research on Geo-information Data Model for Preselected Areas of Geological Disposal of High-level Radioactive Waste

    Science.gov (United States)

    Gao, M.; Huang, S. T.; Wang, P.; Zhao, Y. A.; Wang, H. B.

    2016-11-01

    The geological disposal of high-level radioactive waste (hereinafter referred to "geological disposal") is a long-term, complex, and systematic scientific project, whose data and information resources in the research and development ((hereinafter referred to ”R&D”) process provide the significant support for R&D of geological disposal system, and lay a foundation for the long-term stability and safety assessment of repository site. However, the data related to the research and engineering in the sitting of the geological disposal repositories is more complicated (including multi-source, multi-dimension and changeable), the requirements for the data accuracy and comprehensive application has become much higher than before, which lead to the fact that the data model design of geo-information database for the disposal repository are facing more serious challenges. In the essay, data resources of the pre-selected areas of the repository has been comprehensive controlled and systematic analyzed. According to deeply understanding of the application requirements, the research work has made a solution for the key technical problems including reasonable classification system of multi-source data entity, complex logic relations and effective physical storage structures. The new solution has broken through data classification and conventional spatial data the organization model applied in the traditional industry, realized the data organization and integration with the unit of data entities and spatial relationship, which were independent, holonomic and with application significant features in HLW geological disposal. The reasonable, feasible and flexible data conceptual models, logical models and physical models have been established so as to ensure the effective integration and facilitate application development of multi-source data in pre-selected areas for geological disposal.

  11. Chemical decomposition of high-level nuclear waste storage/disposal glasses under irradiation. 1997 annual progress report

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

    'The objective of this research is to use the sensitive technique of electron spin resonance (ESR) to look for evidence of radiation-induced chemical decomposition of vitreous forms contemplated for immobilization of plutonium and/or high-level nuclear wastes, to interpret this evidence in terms of existing knowledge of glass structure, and to recommend certain materials for further study by other techniques, particularly electron microscopy and measurements of gas evolution by high-vacuum mass spectroscopy. Previous ESR studies had demonstrated that an effect of y rays on a simple binary potassium silicate glass was to induce superoxide (O{sub 2}{sup -}) and ozonide (O{sub 3}{sup -}) as relatively stable product of long-term irradiation Accordingly, some of the first experiments performed as a part of the present effort involved repeating this work. A glass of composition 44 K{sub 2}O: 56 SiO{sub 2} was prepared from reagent grade K{sub 2}CO3 and SiO{sub 2} powders melted in a Pt crucible in air at 1,200 C for 1.5 hr. A sample irradiated to a dose of 1 MGy (1 MGy = 10{sup 8} rad) indeed yielded the same ESR results as before. To test the notion that the complex oxygen ions detected may be harbingers of radiation-induced phase separation or bubble formation, a small-angle neutron scattering (SANS) experiment was performed. SANS is theoretically capable of detecting voids or bubbles as small as 10 \\305 in diameter. A preliminary experiment was carried out with the collaboration of Dr. John Barker (NIST). The SANS spectra for the irradiated and unirradiated samples were indistiguishable. A relatively high incoherent background (probably due to the presence of protons) may obscure scattering from small gas bubbles and therefore decrease the effective resolution of this technique. No further SANS experiments are planned at this time.'

  12. Branch technical position on the use of expert elicitation in the high-level radioactive waste program

    Energy Technology Data Exchange (ETDEWEB)

    Kotra, J.P.; Lee, M.P.; Eisenberg, N.A. [Nuclear Regulatory Commission, Washington, DC (United States); DeWispelare, A.R. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX (United States)

    1996-11-01

    Should the site be found suitable, DOE will apply to the US Nuclear Regulatory Commission for permission to construct and then operate a proposed geologic repository for the disposal of spent nuclear fuel and other high-level radioactive waste at Yucca Mountain. In deciding whether to grant or deny DOE`s license application for a geologic repository, NRC will closely examine the facts and expert judgment set forth in any potential DOE license application. NRC expects that subjective judgments of individual experts and, in some cases, groups of experts, will be used by DOE to interpret data obtained during site characterization and to address the many technical issues and inherent uncertainties associated with predicting the performance of a repository system for thousands of years. NRC has traditionally accepted, for review, expert judgment to evaluate and interpret the factual bases of license applications and is expected to give appropriate consideration to the judgments of DOE`s experts regarding the geologic repository. Such consideration, however, envisions DOE using expert judgments to complement and supplement other sources of scientific and technical information, such as data collection, analyses, and experimentation. In this document, the NRC staff has set forth technical positions that: (1) provide general guidelines on those circumstances that may warrant the use of a formal process for obtaining the judgments of more than one expert (i.e., expert elicitation); and (2) describe acceptable procedures for conducting expert elicitation when formally elicited judgments are used to support a demonstration of compliance with NRC`s geologic disposal regulation, currently set forth in 10 CFR Part 60. 76 refs.

  13. Comparison of mechanical properties of glass-bonded sodalite and borosilicate glass high-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

    O' Holleran, T. P.; DiSanto, T.; Johnson, S. G.; Goff, K. M.

    2000-05-09

    Argonne National Laboratory has developed a glass-bonded sodalite waste form to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The waste form consists of 75 vol.% crystalline sodalite and 25 vol.% glass. Microindentation fracture toughness measurements were performed on this material and borosilicate glass from the Defense Waste Processing Facility using a Vickers indenter. Palmqvist cracking was confined for the glass-bonded sodalite waste form, while median-radial cracking occurred in the borosilicate glass. The elastic modulus was measured by an acoustic technique. Fracture toughness, microhardness, and elastic modulus values are reported for both waste forms.

  14. MATRIX 2 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Raszewski, F; Tommy Edwards, T; David Peeler, D

    2008-10-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt

  15. Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, Dan P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, Matthew S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Luther, Michelle C. [Auburn Univ., AL (United States); Brandenburg, Clayton H. [Univ.of South Carolina, Columbia, SC (United States)

    2016-09-27

    Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

  16. Decay characteristics of once-through LWR and LMFBR spent fuels, high-level wastes, and fuel-assembly structural material wastes

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Alexander, C.W.

    1980-11-01

    The decay characteristics of spent fuel, high-level waste, and fuel-assembly structural material (cladding) waste are presented in the form of ORIGEN2 output tables for (1) a pressurized water reactor operating on a once-through cycle with low-enrichment uranium feed, (2) a boiling-water reactor operating on a once-through cycle with low-enrichment uranium feed, and (3) a liquid-metal fast breeder reactor being fueled with depleted uranium enriched with discharged light water reactor plutonium on a once-through basis. The decay characteristics given include the mass (g), radioactivity (Ci), thermal power (W), photon activity (photons/s and MeV/W-s in 18 energy groups), and neutron activity (neutrons/s) from (..cap alpha..,n) and spontaneous fission events. The first three characteristics are given for each element and for the principal nuclide contributors to the activation products, actinides, and fission products. Also included are a summary description of the ORIGEN2 reactor models that form the basis for the calculated results and a physical description of the fuel assemblies for the three reactors.

  17. MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C; David Diprete, D; Ned Bibler, N

    2008-12-31

    This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

  18. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Burt, D.L.

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection.

  19. Main outcomes from in situ thermo-hydro-mechanical experiments programme to demonstrate feasibility of radioactive high-level waste disposal in the Callovo-Oxfordian claystone

    Directory of Open Access Journals (Sweden)

    G. Armand

    2017-06-01

    Full Text Available In the context of radioactive waste disposal, an underground research laboratory (URL is a facility in which experiments are conducted to demonstrate the feasibility of constructing and operating a radioactive waste disposal facility within a geological formation. The Meuse/Haute-Marne URL is a site-specific facility planned to study the feasibility of a radioactive waste disposal in the Callovo-Oxfordian (COx claystone. The thermo-hydro-mechanical (THM behaviour of the host rock is significant for the design of the underground nuclear waste disposal facility and for its long-term safety. The French National Radioactive Waste Management Agency (Andra has begun a research programme aiming to demonstrate the relevancy of the French high-level waste (HLW concept. This paper presents the programme implemented from small-scale (small diameter boreholes to full-scale demonstration experiments to study the THM effects of the thermal transient on the COx claystone and the strategy implemented in this new programme to demonstrate and optimise current disposal facility components for HLW. It shows that the French high-level waste concept is feasible and working in the COx claystone. It also exhibits that, as for other plastic clay or claystone, heating-induced pore pressure increases and that the THM behaviour is anisotropic.

  20. Regulatory perspectives on model validation in high-level radioactive waste management programs: A joint NRC/SKI white paper

    Energy Technology Data Exchange (ETDEWEB)

    Wingefors, S.; Andersson, J.; Norrby, S. [Swedish Nuclear Power lnspectorate, Stockholm (Sweden). Office of Nuclear Waste Safety; Eisenberg, N.A.; Lee, M.P.; Federline, M.V. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Material Safety and Safeguards; Sagar, B.; Wittmeyer, G.W. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX (United States)

    1999-03-01

    Validation (or confidence building) should be an important aspect of the regulatory uses of mathematical models in the safety assessments of geologic repositories for the disposal of spent nuclear fuel and other high-level radioactive wastes (HLW). A substantial body of literature exists indicating the manner in which scientific validation of models is usually pursued. Because models for a geologic repository performance assessment cannot be tested over the spatial scales of interest and long time periods for which the models will make estimates of performance, the usual avenue for model validation- that is, comparison of model estimates with actual data at the space-time scales of interest- is precluded. Further complicating the model validation process in HLW programs are the uncertainties inherent in describing the geologic complexities of potential disposal sites, and their interactions with the engineered system, with a limited set of generally imprecise data, making it difficult to discriminate between model discrepancy and inadequacy of input data. A successful strategy for model validation, therefore, should attempt to recognize these difficulties, address their resolution, and document the resolution in a careful manner. The end result of validation efforts should be a documented enhancement of confidence in the model to an extent that the model's results can aid in regulatory decision-making. The level of validation needed should be determined by the intended uses of these models, rather than by the ideal of validation of a scientific theory. This white Paper presents a model validation strategy that can be implemented in a regulatory environment. It was prepared jointly by staff members of the U.S. Nuclear Regulatory Commission and the Swedish Nuclear Power Inspectorate-SKI. This document should not be viewed as, and is not intended to be formal guidance or as a staff position on this matter. Rather, based on a review of the literature and previous

  1. Management Plan for the Development of the License Application for a High Level Waste Repository at Yucca Mountain, Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Brocoum, Stephan J.

    1997-01-10

    If the Yucca Mountain Site is recommended and approved for development as a repository, the U.S. Department of Energy (DOE) plans to file a License Application (LA) with the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 60, Disposal of High-Level Radioactive Wastes in Geologic Repositories. The NRC, in accordance with 10 CFR 60, will evaluate DOE's application and adopt, to the extent practicable, the accompanying Environmental Impact Statement (EIS), in connection with the issuance of an authorization to construct the proposed geologic repository. The NRC can issue a license to DOE under 10 CFR 60 only after construction of the geologic repository operations area is substantially complete and the initial LA has been updated in accordance with 10 CFR 60.24. In accordance with the Energy Policy Act of 1992, the NRC is, by rule, to modify its technical requirements and criteria as necessary to be consistent with the U.S. Environmental Protection Agency's new site-specific standards. The NRC staff has proposed to achieve this by drafting a new, separate, site-specific part of the Code of Federal Regulations to be promulgated as 10 CFR 63. While adopting the definitions, administrative, preclosure, retrievability, and quality assurance portions of 10 CFR 60, the NRC staff's proposal for this new part would focus on total-system performance and place no quantitative requirements on the performance of individual subsystems or their components. When the NRC issues a final rule, the DOE will modify its licensing strategy and LA development efforts to conform to the new requirements. Until that time, the applicable repository licensing requirements are those in 10 CFR 60, and the DOE will continue to plan to these requirements. DOE management will, as deemed appropriate, undertake activities to prepare for the eventual issuance of new regulations. The first draft of the LA may be developed to show compliance with the draft regulations prepared

  2. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.

  3. High levels of PAH-metabolites in urine of e-waste recycling workers from Agbogbloshie, Ghana

    Energy Technology Data Exchange (ETDEWEB)

    Feldt, Torsten, E-mail: feldt@bni-hamburg.de [Clinical Research Unit, Bernhard Nocht Institute for Tropical Medicine, Bernhard-Nocht Str. 74, 20359 Hamburg (Germany); Department of Gastroenterology, Hepatology and Infectious Diseases, University Hospital Düsseldorf, Moorenstr. 5, 40225 Düsseldorf (Germany); Fobil, Julius N., E-mail: jfobil@ug.edu.gh [Department of Biological, Environmental and Occupational Health Sciences, School of Public Health, University of Ghana, P.O. Box LG13, Legon (Ghana); Wittsiepe, Jürgen [Department of Hygiene, Social and Environmental Medicine, Ruhr-University Bochum, Universitätsstr. 150, 44801 Bochum (Germany); Wilhelm, Michael, E-mail: wilhelm@hygiene.rub.de [Department of Hygiene, Social and Environmental Medicine, Ruhr-University Bochum, Universitätsstr. 150, 44801 Bochum (Germany); Till, Holger, E-mail: holger.till@giz.de [GIZ — Regional Coordination Unit for HIV and TB (GiZ-ReCHT), 32 Cantonment Crescent, Cantonments, Accra (Ghana); Zoufaly, Alexander [Department of Medicine, Section Infectious Diseases and Tropical Medicine, University Medical Center Hamburg-Eppendorf, Martinistr. 52, 20246 Hamburg (Germany); Burchard, Gerd, E-mail: gerd.burchard@bni-hamburg.de [Clinical Research Unit, Bernhard Nocht Institute for Tropical Medicine, Bernhard-Nocht Str. 74, 20359 Hamburg (Germany); Department of Medicine, Section Infectious Diseases and Tropical Medicine, University Medical Center Hamburg-Eppendorf, Martinistr. 52, 20246 Hamburg (Germany); Göen, Thomas, E-mail: thomas.goeen@ipasum.med.uni-erlangen.de [Institute and Outpatient Clinic of Occupational, Social and Environmental Medicine, University of Erlangen-Nuremberg, Schillerstr. 25/29, 91054 Erlangen (Germany)

    2014-01-01

    The informal recycling of electronic waste (e-waste) is an emerging source of environmental pollution in Africa. Among other toxins, polycyclic aromatic hydrocarbons (PAHs) are a major health concern for exposed individuals. In a cross-sectional study, the levels of PAH metabolites in the urine of individuals working on one of the largest e-waste recycling sites of Africa, and in controls from a suburb of Accra without direct exposure to e-waste recycling activities, were investigated. Socioeconomic data, basic health data and urine samples were collected from 72 exposed individuals and 40 controls. In the urine samples, concentrations of the hydroxylate PAH metabolites (OH-PAH) 1-hydroxyphenanthrene (1-OH-phenanthrene), the sum of 2- and 9-hydroxyphenanthrene (2-/9-OH-phenanthrene), 3-hydroxyphenanthrene (3-OH-phenanthrene), 4-hydroxyphenanthrene (4-OH-phenanthrene) and 1-hydroxypyrene (1-OH-pyrene), as well as cotinine and creatinine, were determined. In the exposed group, median urinary concentrations were 0.85 μg/g creatinine for 1-OH-phenanthrene, 0.54 μg/g creatinine for 2-/9-OH-phenanthrene, 0.99 μg/g creatinine for 3-OH-phenanthrene, 0.22 μg/g creatinine for 4-OH-phenanthrene, and 1.33 μg/g creatinine for 1-OH-pyrene, all being significantly higher compared to the control group (0.55, 0.37, 0.63, 0.11 and 0.54 μg/g creatinine, respectively). Using a multivariate linear regression analysis including sex, cotinine and tobacco smoking as covariates, exposure to e-waste recycling activities was the most important determinant for PAH exposure. On physical examination, pathological findings were rare, but about two thirds of exposed individuals complained about cough, and one quarter about chest pain. In conclusion, we observed significantly higher urinary PAH metabolite concentrations in individuals who were exposed to e-waste recycling compared to controls who were not exposed to e-waste recycling activities. The impact of e-waste recycling on exposure to

  4. Scoring methods and results for qualitative evaluation of public health impacts from the Hanford high-level waste tanks. Integrated Risk Assessment Program

    Energy Technology Data Exchange (ETDEWEB)

    Buck, J.W.; Gelston, G.M.; Farris, W.T.

    1995-09-01

    The objective of this analysis is to qualitatively rank the Hanford Site high-level waste (HLW) tanks according to their potential public health impacts through various (groundwater, surface water, and atmospheric) exposure pathways. Data from all 149 single-shell tanks (SSTs) and 23 of the 28 double-shell tanks (DSTs) in the Tank Waste Remediation System (TWRS) Program were analyzed for chemical and radiological carcinogenic as well as chemical noncarcinogenic health impacts. The preliminary aggregate score (PAS) ranking system was used to generate information from various release scenarios. Results based on the PAS ranking values should be considered relative health impacts rather than absolute risk values.

  5. Inductively coupled plasma-mass spectrometry studies of the chemistry of fission products and actinides in high level wastes: lessons that can be applied to environmental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Kinard, W.F. [Dept. of Chemistry, Coll. of Charleston, SC (United States); Bibler, N.E. [Westinghouse Savannah River Technology Center (SRTC), Westinghouse Savannah River Corp., Aiken, SC (United States); Coleman, C.J. [Westinghouse Savannah River Technology Center (SRTC), Westinghouse Savannah River Corp., Aiken, SC (United States); Wyrick, S.B. [Science Applications International, Gaithersburg, MD (United States)

    1994-12-31

    Actinide and fission product concentrations in HLW (high level wastes) from the Savannah River Site (SRS) in South Carolina and from Tank 101-SY at the Hanfrod Site in the state of Washington have been measured by inductively coupled plasma-mass spectrometry. Isotopic assignments based on fission yield predictions has enabled the analyses to be made without further separations other than the chemical processing used to separate waste streams. Isotopic patterns related to weapons reactor procuction are proposed as possible tracers for environmental measurements. (orig.)

  6. New Technological Options to Manage High Level Waste; Nuevas tecnologias para gestionar resiudos radiactivos de alta actividad

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Romero, E. M.

    2007-07-01

    Nuclear energy renaissance and its expansion in time and space has renewed the need for minimization technologies applicable to nuclear wastes. The minimization technologies include new power reactor concepts, Generation IV, and dedicated technologies like Partitioning and Transmutation of the actinides contained in the spent fuel. These technologies apply the principle of classification and recycling to the spent fuel to transform what at present is an environmental hazard into an energy source. the waste minimization technologies are also relevant for countries planning the reduction or phase-out of nuclear energy, as they will allow minimizing the size and number of the final waste repositories. Present estimations indicate that reductions as large as a factor 100 in the amount (radiotoxicity) of long lived nuclear waste are feasibly, with a modest increase on the final electricity cost. (Author)

  7. Thermal-Hydrology Simulations of Disposal of High-Level Radioactive Waste in a Single Deep Borehole

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Freeze, Geoffrey A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    Simulations of thermal-hydrology were carried out for the emplacement of spent nuclear fuel canisters and cesium and strontium capsules using the PFLOTRAN simulator. For the cesium and strontium capsules the analysis looked at disposal options such as different disposal configurations and surface aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperature increases, and fluxes at the top of the disposal zone. Simulations of cesium and strontium capsule disposal predict that surface aging and/or emplacement of the waste at the top of the disposal zone reduces thermal effects and vertical fluid fluxes. Smaller waste canisters emplaced over a longer disposal zone create the smallest thermal effect and vertical fluid fluxes no matter the age of the waste or depth of emplacement.

  8. Fall 2010 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post Closure Permit for the INTEC Waste Calcining Facility and the CPP 601/627/640 Facility at the INL Site

    Energy Technology Data Exchange (ETDEWEB)

    Boehmer, Ann

    2010-11-01

    The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under an approved Hazardous Waste Management Act/Resource Conservation and Recovery Act (HWMA/RCRA) Closure Plan. Vessels and spaces were grouted and then covered with a concrete cap. The Idaho Department of Environmental Quality issued a final HWMA/RCRA post-closure permit on September 15, 2003, with an effective date of October 16, 2003. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment. The post closure permit also includes semiannual reporting requirements under Permit Conditions III.H. and I.U. These reporting requirements have been combined into this single semiannual report, as agreed between the Idaho Cleanup Project and Idaho Department of Environmental Quality. The Permit Condition III.H. portion of this report includes a description and the results of field methods associated with groundwater monitoring of the Waste Calcining Facility. Analytical results from groundwater sampling, results of inspections and maintenance of monitoring wells in the Waste Calcining Facility groundwater monitoring network, and results of inspections of the concrete cap are summarized. The Permit Condition I.U. portion of this report includes noncompliances not otherwise required to be reported under Permit Condition I.R. (advance notice of planned changes to facility activity which may result in a noncompliance) or Permit Condition I.T. (reporting of noncompliances which may endanger human health or the environment). This report also provides groundwater sampling results for wells that were installed and monitored as part of the Phase 1 post-closure period of the landfill closure components in accordance with HWMA/RCRA Landfill Closure Plan for the CPP-601 Deep

  9. Shielding calculations with SCALE/MAVRIC and comparison with measurements for the TN85 cask with vitrified high level radioactive waste

    Science.gov (United States)

    Thiele, Holger; Börst, Frank-Michael

    2017-09-01

    A series of dose rate/spectra measurements in the German interim storage facility Gorleben was carried out at a TN85 cask in April 2009. This type of cask is used for the transport and interim storage of vitrified high level radioactive waste (HAW) from reprocessing. The aim of this work is to assess the shielding component MAVRIC of the SCALE code system with these measurements for the use in the German Bundesamt für Kerntechnische Entsorgungssicherheit (BfE).

  10. Basic investigation and analysis for preferred host rocks and natural analogue study area with reference to high level radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Ryul; Park, J. K.; Hwang, D. H.; Lee, J. H.; Yun, H. S.; Kim, D. Y.; Park, H. S.; Koo, S. B.; Cho, J. D.; Kim, K. E. [Korea Inst. of Geology, Mining and Materials, Taejon (Korea, Republic of)

    1997-12-01

    The purpose of this study is basic investigation and analysis for preferred host rocks and natural analogue study area to develope underground disposal technique of high level radioactive waste in future. The study has been done for the crystalline rocks(especially granitic rocks) with emphasis of abandoned metallic mines and uranium ore deposits, and for the geological structure study by using gravity and aeromagnetic data. 138 refs., 54 tabs., 130 figs. (author)

  11. A report on high-level nuclear waste transportation: Prepared pursuant to assembly concurrent resolution No. 8 of the 1987 Nevada Legislature

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-01

    This report has been prepared by the staff of the State of Nevada Agency for Nuclear Projects/Nuclear Waste Project Office (NWPO) in response to Assembly Concurrent Resolution No. 8 (ACR 8), passed by the Nevada State Legislature in 1987. ACR 8 directed the NWPO, in cooperation with affected local governments and the Legislative committee on High-Level Radioactive Waste, to prepare this report which scrutinizes the US Department of Energy`s (DOE) plans for transportation of high-level radioactive waste to the proposed yucca Mountain repository, which reviews the regulatory structure under which shipments to a repository would be made and which presents NWPO`s plans for addressing high-level radioactive waste transportation issues. The report is divided into three major sections. Section 1.0 provides a review of DOE`s statutory requirements, its repository transportation program and plans, the major policy, programmatic, technical and institutional issues and specific areas of concern for the State of Nevada. Section 2.0 contains a description of the current federal, state and tribal transportation regulatory environment within which nuclear waste is shipped and a discussion of regulatory issues which must be resolved in order for the State to minimize risks and adverse impacts to its citizens. Section 3.0 contains the NWPO plan for the study and management of repository-related transportation. The plan addresses four areas, including policy and program management, regulatory studies, technical reviews and studies and institutional relationships. A fourth section provides recommendations for consideration by State and local officials which would assist the State in meeting the objectives of the plan.

  12. High levels of PAH-metabolites in urine of e-waste recycling workers from Agbogbloshie, Ghana.

    Science.gov (United States)

    Feldt, Torsten; Fobil, Julius N; Wittsiepe, Jürgen; Wilhelm, Michael; Till, Holger; Zoufaly, Alexander; Burchard, Gerd; Göen, Thomas

    2014-01-01

    The informal recycling of electronic waste (e-waste) is an emerging source of environmental pollution in Africa. Among other toxins, polycyclic aromatic hydrocarbons (PAHs) are a major health concern for exposed individuals. In a cross-sectional study, the levels of PAH metabolites in the urine of individuals working on one of the largest e-waste recycling sites of Africa, and in controls from a suburb of Accra without direct exposure to e-waste recycling activities, were investigated. Socioeconomic data, basic health data and urine samples were collected from 72 exposed individuals and 40 controls. In the urine samples, concentrations of the hydroxylate PAH metabolites (OH-PAH) 1-hydroxyphenanthrene (1-OH-phenanthrene), the sum of 2- and 9-hydroxyphenanthrene (2-/9-OH-phenanthrene), 3-hydroxyphenanthrene (3-OH-phenanthrene), 4-hydroxyphenanthrene (4-OH-phenanthrene) and 1-hydroxypyrene (1-OH-pyrene), as well as cotinine and creatinine, were determined. In the exposed group, median urinary concentrations were 0.85 μg/g creatinine for 1-OH-phenanthrene, 0.54 μg/g creatinine for 2-/9-OH-phenanthrene, 0.99 μg/g creatinine for 3-OH-phenanthrene, 0.22 μg/g creatinine for 4-OH-phenanthrene, and 1.33 μg/g creatinine for 1-OH-pyrene, all being significantly higher compared to the control group (0.55, 0.37, 0.63, 0.11 and 0.54 μg/g creatinine, respectively). Using a multivariate linear regression analysis including sex, cotinine and tobacco smoking as covariates, exposure to e-waste recycling activities was the most important determinant for PAH exposure. On physical examination, pathological findings were rare, but about two thirds of exposed individuals complained about cough, and one quarter about chest pain. In conclusion, we observed significantly higher urinary PAH metabolite concentrations in individuals who were exposed to e-waste recycling compared to controls who were not exposed to e-waste recycling activities. The impact of e-waste recycling on exposure to

  13. Risk-based systems analysis of emerging high-level waste tank remediation technologies. Volume 2: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Peters, B.B.; Cameron, R.J.; McCormack, W.D. [Enserch Environmental Corp., Richland, WA (United States)

    1994-08-01

    The objective of DOE`s Radioactive Waste Tank Remediation Technology Focus Area is to identify and develop new technologies that will reduce the risk and/or cost of remediating DOE underground waste storage tanks and tank contents. There are, however, many more technology investment opportunities than the current budget can support. Current technology development selection methods evaluate new technologies in isolation from other components of an overall tank waste remediation system. This report describes a System Analysis Model developed under the US Department of Energy (DOE) Office of Technology Development (OTD) Underground Storage Tank-Integrated Demonstration (UST-ID) program. The report identifies the project objectives and provides a description of the model. Development of the first ``demonstration`` version of this model and a trial application have been completed and the results are presented. This model will continue to evolve as it undergoes additional user review and testing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zyryanov, V.N. [Argonne National Lab., IL (United States); Vance, E.R. [ANSTO, Menai (Australia). Materials Division

    1996-12-31

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

  15. Japan-Australia co-operative program on research and development of technology for the management of high level radioactive wastes. Final report 1985 to 1998

    Energy Technology Data Exchange (ETDEWEB)

    Hart, K.; Vance, E.; Lumpkin, G. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia); Mitamura, H.; Banba, T. [Japan Atomic Energy Research Inst. Tokai, Ibaraki (Japan)

    1998-12-01

    The overall aim of the Co-operative Program has been to promote the exchange of information on technology for the management of High-Level Wastes (HLW) and to encourage research and development relevant to such technology. During the 13 years that the Program has been carried out, HLW management strategies have matured and developed internationally, and Japan has commenced construction of a domestic reprocessing and vitrification facility for HLW. The HLW management strategy preferred is a national decision. Many countries are using vitrification, direct disposal of spent fuel or a combination of both to handle their existing wastes whereas others have deferred the decision. The work carried out in the Co-operative Program provides strong scientific evidence that the durability of ceramic waste forms is not significantly affected by radiation damage and that high loadings of actinide elements can be incorporated into specially designed ceramic waste forms. Moreover, natural minerals have been shown to remain as closed systems for U and Th for up to 2.5 b y. All of these results give confidence in the ability of second generation waste forms, such as Synroc, to handle future waste arisings that may not be suitable for vitrification 87 refs., 15 tabs., 22 figs.

  16. FURTHER DEVELOPMENT OF MODIFIED MONOSODIUM TITANATE, AN IMPROVED SORBENT FOR PRETREATMENT OF HIGH LEVEL NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K.; Hobbs, D.; Fondeur, F.; Fink, S.

    2011-01-12

    High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include caustic side solvent extraction, for Cs-137 removal, and sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST). The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239, and Pu-240. This paper describes recent results from the development of an improved titanate material that exhibits increased removal kinetics and effective capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material.

  17. Some technical and legal problems relating to the storage of high-level radioactive waste and the use of nuclear power sources on space satellites

    Energy Technology Data Exchange (ETDEWEB)

    Herkommer, E.; Wollenschlaeger, M.

    1985-04-01

    A brief survey is presented summarizing the main characteristics of radioactive wastes and the various waste management strategies. Subsequently, the technical and legal problems encountered with the final disposal of high-level radioactive waste and with the use of nuclear power sources on space satellites are reviewed. It is shown that both in terms of technology and law, a sound basis is already available upon which the problem of HAW disposal in space can be tackled. On the legal level, however, existing norms and regulations need to be supplemented and improved by more concrete provisions, and this task should be started now. An international agreement concerning HAW management in space is said to be indispensable.

  18. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.W.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-09-22

    One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys.

  19. Summary report of first and foreign high-level waste repository concepts; Technical report, working draft 001

    Energy Technology Data Exchange (ETDEWEB)

    Hanke, P.M.

    1987-11-04

    Reference repository concepts designs adopted by domestic and foreign waste disposal programs are reviewed. Designs fall into three basic categories: deep borehole from the surface; disposal in boreholes drilled from underground excavations; and disposal in horizontal tunnels or drifts. The repository concepts developed in Sweden, Switzerland, Finland, Canada, France, Japan, United Kingdom, Belgium, Italy, Holland, Denmark, West Germany and the United States are described. 140 refs., 315 figs., 19 tabs.

  20. SPRAY CALCINATION REACTOR

    Science.gov (United States)

    Johnson, B.M.

    1963-08-20

    A spray calcination reactor for calcining reprocessin- g waste solutions is described. Coaxial within the outer shell of the reactor is a shorter inner shell having heated walls and with open regions above and below. When the solution is sprayed into the irner shell droplets are entrained by a current of gas that moves downwardly within the inner shell and upwardly between it and the outer shell, and while thus being circulated the droplets are calcined to solids, whlch drop to the bottom without being deposited on the walls. (AEC) H03 H0233412 The average molecular weights of four diallyl phthalate polymer samples extruded from the experimental rheometer were redetermined using the vapor phase osmometer. An amine curing agent is required for obtaining suitable silver- filled epoxy-bonded conductive adhesives. When the curing agent was modified with a 47% polyurethane resin, its effectiveness was hampered. Neither silver nor nickel filler impart a high electrical conductivity to Adiprenebased adhesives. Silver filler was found to perform well in Dow-Corning A-4000 adhesive. Two cascaded hot-wire columns are being used to remove heavy gaseous impurities from methane. This purified gas is being enriched in the concentric tube unit to approximately 20% carbon-13. Studies to count low-level krypton-85 in xenon are continuing. The parameters of the counting technique are being determined. The bismuth isotopes produced in bismuth irradiated for polonium production are being determined. Preliminary data indicate the presence of bismuth207 and bismuth-210m. The light bismuth isotopes are probably produced by (n,xn) reactions bismuth-209. The separation of uranium-234 from plutonium-238 solutions was demonstrated. The bulk of the plutonium is removed by anion exchange, and the remainder is extracted from the uranium by solvent extraction techniques. About 99% of the plutonium can be removed in each thenoyltrifluoroacetone extraction. The viscosity, liquid density, and

  1. Applications of inductively coupled plasma-mass spectrometry to the determination of actinides and fission products in high level radioactive wastes at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Kinard, W.F. [College of Charleston, SC (United States). Dept. of Chemistry; Bibler, N.E.; Coleman, C.J.; Dewberry, R.A.; Boyce, W.T. [Westinghouse Savannah River Technology Center, Aiken, SC (United States); Wyrick, S.B. [Science Applications International, Gaithersburg, MD (United States)

    1995-12-31

    Four years of experience in applying inductively coupled plasma-mass spectrometry (ICP-MS) to the analysis of actinides and fission products in high level waste (HLW) samples at the Savannah River Site has led to the development of a number of techniques to aid in the interpretation of the mass spectral data. The goal has been to develop rapid and reliable analytical procedures that provide the necessary chemical and isotopic information to answer the process needs of the customers. Techniques that have been developed include the writing of computer software to strip the experimental data from the instrumental data files into spreadsheets or into a spectral data processing package so that the raw mass spectra can be overlain for comparison or plotted with higher output resolution. These procedures have been applied to problems ranging from the analysis of the high level waste tanks to reactor moderator water as well as environmental samples. Criticality safety analyses in some HLW waste treatment processes depend upon actinide concentration and isotopic information generated by ICP-MS, particularly in tanks with high concentrations of {sup 137}Cs and {sup 90}Sr. Experimental results for a number of these applications will be presented. These procedures represent a considerable saving in time and expense as compared to conventional chemical separation followed by radiochemical analyses, as well as decreased radiation exposure for the analysts.

  2. Is Yucca Mountain a long-term solution for disposing of US spent nuclear fuel and high-level radioactive waste?

    Science.gov (United States)

    Thorne, M C

    2012-06-01

    On 26 January 2012, the Blue Ribbon Commission on America's Nuclear Future released a report addressing, amongst other matters, options for the managing and disposal of high-level waste and spent fuel. The Blue Ribbon Commission was not chartered as a siting commission. Accordingly, it did not evaluate Yucca Mountain or any other location as a potential site for the storage or disposal of spent nuclear fuel and high-level waste. Nevertheless, if the Commission's recommendations are followed, it is clear that any future proposals to develop a repository at Yucca Mountain would require an extended period of consultation with local communities, tribes and the State of Nevada. Furthermore, there would be a need to develop generally applicable regulations for disposal of spent fuel and high-level radioactive waste, so that the Yucca Mountain site could be properly compared with alternative sites that would be expected to be identified in the initial phase of the site-selection process. Based on what is now known of the conditions existing at Yucca Mountain and the large number of safety, environmental and legal issues that have been raised in relation to the DOE Licence Application, it is suggested that it would be imprudent to include Yucca Mountain in a list of candidate sites for future evaluation in a consent-based process for site selection. Even if there were a desire at the local, tribal and state levels to act as hosts for such a repository, there would be enormous difficulties in attempting to develop an adequate post-closure safety case for such a facility, and in showing why this unsaturated environment should be preferred over other geological contexts that exist in the USA and that are more akin to those being studied and developed in other countries.

  3. CASTOR {sup ®} and CONSTOR {sup ®}. A well established system for the dry storage of spent fuel and high level waste

    Energy Technology Data Exchange (ETDEWEB)

    Wimmer, Hannes; Skrzyppek, Juergen; Koebl, Michael [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2015-06-01

    The German company GNS Gesellschaft fuer Nuklear-Service mbH today looks back on more than 30 years of operational experience with dual-purpose casks for the transport and storage of spent nuclear fuel (SNF) from nuclear power plants and high level waste (HLW) from reprocessing. Following customer demands, GNS developed two different cask types for SNF. By now, almost 1,300 GNS-casks are in operation worldwide. This article gives an overview over several national and international projects and shows the bandwidth of customised solutions by GNS.

  4. Implications of theories of asteroid and comet impact for policy options for management of spent nuclear fuel and high-level radioactive wastes

    Science.gov (United States)

    Trask, Newell J.

    1994-01-01

    Concern with the threat posed by terrestrial asteroid and comet impacts has heightened as the catastrophic consequences of such events have become better appreciated. Although the probabilities of such impacts are very small, a reasonable question for debate is whether such phenomena should be taken into account in deciding policy for the management of spent fuel and high-level radioactive waste. The rate at which asteroid or comet impacts would affect areas of surface storage of radioactive waste is about the same as the estimated rate at which volcanic activity would affect the Yucca Mountain area. The Underground Retrievable Storage (URS) concept could satisfactorily reduce the risk from cosmic impact with its associated uncertainties in addition to providing other benefits described by previous authors.

  5. Toward a risk assessment of the spent fuel and high-level nuclear waste disposal system. Risk assessment requirements, literature review, methods evaluation: an interim report

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, L.D.; Hill, D.; Rowe, M.D.; Stern, E.

    1986-04-01

    This report provides background information for a risk assessment of the disposal system for spent nuclear fuel and high-level radioactive waste (HLW). It contains a literature review, a survey of the statutory requirements for risk assessment, and a preliminary evaluation of methods. The literature review outlines the state of knowledge of risk assessment and accident consequence analysis in the nuclear fuel cycle and its applicability to spent fuel and HLW disposal. The survey of statutory requirements determines the extent to which risk assessment may be needed in development of the waste-disposal system. The evaluation of methods reviews and evaluates merits and applicabilities of alternative methods for assessing risks and relates them to the problems of spent fuel and HLW disposal. 99 refs.

  6. Sensitivity analysis on mechanical stability of the underground excavations for an high-level radioactive waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Hwa; Kwon, Sang Ki; Choi, Jong Won; Kang, Chul Hyung

    2001-01-01

    For the safe design of an underground nuclear waste repository, it is necessary to investigate the influence of the major parameters on the tunnel stability. In this study, sensitivity analysis was carried out to find the major parameters on the mechanical stability point of view. Fourteen parameters consisted of 10 site parameters and 4 design parameters were included in the FLAC3D. From the numerical analyses employing single parameter variation, it was possible to determine important parameters. In order to investigate the interaction between the parameters, fractional factorial design for the parameters, such as in situ stress ratio, depth, tunnel dimensions, joint spacing, joint stiffness, friction angle, and rock strength, was carried out. And in order to investigate the interaction between design parameters, fractional factorial design for parameters, such as in situ stress, depth, tunnel size, tunnel spacing and borehole spacing, was carried out.

  7. AN EVALUATION OF HYDROGEN INDUCED CRACKING SUSCEPTIBILITY OF TITANIUM ALLOYS IN US HIGH-LEVEL NUCLEAR WASTE REPOSITORY ENVIRONMENTS

    Energy Technology Data Exchange (ETDEWEB)

    G. De; K. Mon; G. Gordon; D. Shoesmith; F. Hua

    2006-02-21

    This paper evaluates hydrogen-induced cracking (HIC) susceptibility of titanium alloys in environments anticipated in the Yucca Mountain nuclear waste repository with particular emphasis on the. effect of the oxide passive film on the hydrogen absorption process of titanium alloys being evaluated. The titanium alloys considered in this review include Ti 2, 5 , 7, 9, 11, 12, 16, 17, 18, 24 and 29. In general, the concentration of hydrogen in a titanium alloy can increase due to absorption of atomic hydrogen produced from passive general corrosion of that alloy or galvanic coupling of it to a less noble metal. It is concluded that under the exposure conditions anticipated in the Yucca Mountain repository, the HIC of titanium drip shield will not occur because there will not be sufficient hydrogen in the metal even after 10,000 years of emplacement. Due to the conservatisms adopted in the current evaluation, this assessment is considered very conservative.

  8. The challenge of long-term participatory repository governance. Lessons learned for high level radioactive waste and spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Landstroem, Catharina [East Anglia Univ., Norwich (United Kingdom). School of Environmental Sciences; Barbier, Jan-Willem [Antwerp Univ. (Belgium)

    2012-12-15

    Voluntaristic siting procedures for deep geological repositories are becoming increasingly common; they reconfigure the relationship of repositories and society in ways that have implications for the long-term governance of these facilities. This paper identifies three challenges emerging in relation to this question: principles of monitoring, repository content, and facility closure. This paper discusses them in a comparison with similar challenges being addressed in Belgian partnerships founded to facilitate the siting and design of a low- and intermediate level short lived waste repository. The empirical exploration confirms the importance of securing stakeholder engagement throughout the repository lifecycle, for which there is a need to develop knowledge about how to encourage long-term democratic governance systems.

  9. End of FY10 report - used fuel disposition technical bases and lessons learned : legal and regulatory framework for high-level waste disposition in the United States.

    Energy Technology Data Exchange (ETDEWEB)

    Weiner, Ruth F.; Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Rechard, Robert Paul; Perry, Frank (Los Alamos National Laboratory, Los Alamos, NM); Jenkins-Smith, Hank C. (University of Oklahoma, Norman, OK); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Nutt, Mark (Argonne National Laboratory, Argonne, IL); Cotton, Tom (Complex Systems Group, Washington DC)

    2010-09-01

    This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

  10. Enrichment Of Calcined Magnesite

    OpenAIRE

    GENCE, Nermin

    2017-01-01

    In this study, the possibility ofconcentration of magnesite are which contains high dolomite, low iron andsiliceous gangue was investigated by heavy medium separation, shaking table and vashing-screeningafter calcination.

  11. Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

    Energy Technology Data Exchange (ETDEWEB)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs.

  12. Chemical speciation of strontium, americium, and curium in high level waste: Predictive modeling of phase partitioning during tank processing. Annual progress report, October 1996--September 1997

    Energy Technology Data Exchange (ETDEWEB)

    Felmy, A.R. [Pacific Northwest National Lab., Richland, WA (US); Choppin, G. [Florida State Univ., Tallahassee, FL (US)

    1997-12-31

    'The program at Florida State University was funded to collaborate with Dr. A. Felmy (PNNL) on speciation in high level wastes and with Dr. D. Rai (PNNL) on redox of Pu under high level waste conditions. The funding provided support for 3 research associates (postdoctoral researchers) under Professor G. R. Choppin as P.I. Dr. Kath Morris from U. Manchester (Great Britain), Dr. Dean Peterman and Dr. Amy Irwin (both from U. Cincinnati) joined the laboratory in the latter part of 1996. After an initial training period to become familiar with basic actinide chemistry and radiochemical techniques, they began their research. Dr. Peterman was assigned the task of measuring Th-EDTA complexation prior to measuring Pu(IV)-EDTA complexation. These studies are associated with the speciation program with Dr. Felmy. Drs. Morris and Irwin initiated research on redox of plutonium with agents present in the Hanford Tanks as a result of radiolysis or from use in separations. The preliminary results obtained thus far are described in this report. It is expected that the rate of progress will continue to increase significantly as the researchers gain more experience with plutonium chemistry.'

  13. Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Part 1. Geological environment of Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the geological disposal concept for high-level radioactive wastes (HLW) in Japan. The present document, the part 1 of the progress report, describes first in detail the role of geological environment in high-level radioactive wastes disposal, the features of Japanese geological environment, and programs to proceed the investigation in geological environment. The following chapter summarizes scientific basis for possible existence of stable geological environment, stable for a long period needed for the HLW disposal in Japan including such natural phenomena as volcano and faults. The results of the investigation of the characteristics of bed-rocks and groundwater are presented. These are important for multiple barrier system construction of deep geological disposal. The report furthermore describes the present status of technical and methodological progress in investigating geological environment and finally on the results of natural analog study in Tono uranium deposits area. (Ohno, S.)

  14. Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects; Report for October 1987--June 1989

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.E.; Arnold, W.D.; O`Kelley, G.D.; Case, F.I.; Land, J.F.

    1989-08-01

    Information that is being developed by projects within the Department of Energy (DOE) pertinent to the potential geochemical behavior of radionuclides at candidate sites for a high-level radioactive waste repository is being evaluated by Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC). During this report period, all experiments were conducted with tuff from the proposed high-level nuclear waste site at Yucca Mountain, Nevada. The principal emphasis in this report period was on column studies of migration of uranium and technetium in water from well J-13 at the Yucca Mountain site. Columns 1 cm in diameter and about 5 cm long were constructed and carefully packed with ground tuff. The characteristics of the columns were tested by determination of elution curves of tritium and TcO{sub 4}{sup {minus}}. Elution peaks obtained in past studies with uranium were asymmetrical and the shapes were often complex, observations that suggested irreversibilities in the sorption reaction. To try to understand these observations, the effects of flow rate and temperature on uranium migration were studied in detail. Sorption ratios calculated from the elution peaks became larger as the flow rate decreased and as the temperature increased. These observations support the conclusion that the sorption of uranium is kinetically hindered. To confirm this, batch sorption ratio experiments were completed for uranium as a function of time for a variety of conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Levy, Salomon, E-mail: slevy112@aol.co [3425 South Bascom Avenue, Suite 225, Campbell, CA 95008 (United States)

    2010-10-15

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

  16. Social scientist on board in long-term management of high level and/or long-lived radioactive waste in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Parotte, C. [Spiral Research Center, Department of Political Sciences, Faculty of Law, University of Liege (Belgium)

    2013-07-01

    In Belgium, the long-term management of radioactive waste is under the exclusive competence of the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (knew as ONDRAF/NIRAS). Unlike low-level waste, no institutional policy has yet been formally approved for the long-term management of high level and/or long-lived radioactive waste (knew as B and C waste). In this context, ONDRAF/NIRAS considers the public and stakeholders' participation as an essential factor in the formulation of an effective and legitimate policy. This is why it has decided to integrate them in different ways during the elaboration of the Waste Plan (ONDRAF/NIRAS-document containing guidelines to make a principled policy decision about nuclear waste management). To do so, social scientists have been regularly mobilized either as external evaluators, follow-up committee members, or participatory observants. Hence, the Waste Plan is only the first step in a long decision-making process. For a PhD student under contract with ONDRAF/NIRAS, this mandate consists of thinking out a way to construct an inter-organizational innovative communication system that would be participative, transparent and embedded in a long-term perspective, thus integrating all the further legal steps to take throughout the decision-making process. In this regard, two paradoxical constraints must be taken into account: on the one hand, my own influence on the legal decision-making process should remain limited, because of a series of constraints, lock-ins and previous decisions which have to be respected; on the other hand, ONDRAF/NIRAS expects the research conclusions to be policy relevant and useful. In this paper, the purpose is twofold. Firstly, the issues raised by this policy mandate is an opportunity to question the per-formative dimensions of the social scientist in the decision-making process and, more specifically, to have a reflexive view on our position as PhD Student. Secondly, assuming the

  17. Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock-salt formations

    Energy Technology Data Exchange (ETDEWEB)

    Smailos, E.; Schwarzkopf, W.; Koester, R.; Fiehn, B.; Halm, G. [Kernforschungszentrum Karlsruhe GmbH (DE)

    1991-12-31

    In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. Investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have also been carried out in order to complete the results available to date. The three steels (fine-grained steel, low-carbon steel, cast steel) investigated and Ti 99.8-Pd resisted pitting and crevice corrosion as well as stress-corrosion cracking under all test conditions. Gamma dose-rates of 1 Gy/h - 100 Gy/h or H{sub 2}S concentrations in the brines as well as welding and explosion plating did not influence noticeably the corrosion behaviour of the materials. Furthermore, the determined corrosion rates of the steels (50 {mu}m/a-250 {mu}m/a, depending on the test conditions) are intercomparable and imply technically acceptable corrosion allowances for the thick-walled containers discussed. For Ti 99.8-Pd no detectable corrosion was observed. By contrast, Hastelloy C4 proved susceptible to pitting and crevice corrosion at gamme dose-rates higher than 1 Gy/h and in the presence of H{sub 2}S (25 mg/l) in Q-brine. The materials Ni Resist D2 and D4, cast iron and Si-cast iron corroded at negligible rates in the in-situ experiments performed in rock salt/limited amounts of NaCI-brine. Nevertheless, these materials must be ruled out as container materials because they have proved to be susceptible to pitting and intergranular corrosion in previous laboratory studies conducted with MgCI{sub 2}-rich brine (Q-brine) in excess. 15 refs.; 29 figs.; 7 tabs.

  18. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  19. Environmental Assessment for the Closure of the High-Level Waste Tanks in F- & H-Areas at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    N/A

    1996-07-31

    This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) to assess the potential environmental impacts associated with the closure of 51 high-level radioactive waste tanks and tank farm ancillary equipment (including transfer lines, evaporators, filters, pumps, etc) at the Savannah River Site (SRS) located near Aiken, South Carolina. The waste tanks are located in the F- and H-Areas of SRS and vary in capacity from 2,839,059 liters (750,000 gallons) to 4,921,035 liters (1,300,000 gallons). These in-ground tanks are surrounded by soil to provide shielding. The F- and H-Area High-Level Waste Tanks are operated under the authority of Industrial Wastewater Permits No.17,424-IW; No.14520, and No.14338 issued by the South Carolina Department of Health and Environmental Control (SCDHEC). In accordance with the Permit requirements, DOE has prepared a Closure Plan (DOE, 1996) and submitted it to SCDHEC for approval. The Closure Plan identifies all applicable or relevant and appropriate regulations, statutes, and DOE Orders for closing systems operated under the Industrial Wastewater Permits. When approved by SCDHEC, the Closure Plan will present the regulatory process for closing all of the F- and H-Area High Level Waste Tanks. The Closure Plan establishes performance objectives or criteria to be met prior to closing any tank, group of tanks, or ancillary tank farm equipment. The proposed action is to remove the residual wastes from the tanks and to fill the tanks with a material to prevent future collapse and bind up residual waste, to lower human health risks, and to increase safety in and around the tanks. If required, an engineered cap consisting of clay, backfill (soil), and vegetation as the final layer to prevent erosion would be applied over the tanks. The selection of tank system closure method will be evaluated against the following Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) criteria described in 40

  20. FEBEX project: full-scale engineered barriers experiment for a deep geological repository for high level radioactive waste in crystalline host rock

    Energy Technology Data Exchange (ETDEWEB)

    Alberid, J.; Barcala, J. M.; Campos, R.; Cuevas, A. M.; Fernandez, E. [Ciemat. Madrid (Spain)

    2000-07-01

    FEBEX has the multiple objective of demonstrating the feasibility of manufacturing, handling and constructing the engineered barriers and of developing codes for the thermo-hydro-mechanical and thermo-hydro-geochemical performance assessment of a deep geological repository for high level radioactive wastes. These objectives require integrated theoretical and experimental development work. The experimental work consists of three parts: an in situ test, a mock-up test and a series of laboratory tests. The experiments is based on the Spanish reference concept for crystalline rock, in which the waste capsules are placed horizontally in drifts surround by high density compacted bentonite blocks. In the two large-scale tests, the thermal effects of the wastes were simulated by means of heaters; hydration was natural in the in situ test and controlled in the mock-up test. The large-scale tests, with their monitoring systems, have been in operation for more than two years. the demonstration has been achieved in the in situ test and there are great expectation that numerical models sufficiently validated for the near-field performance assessment will be achieved. (Author)

  1. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P.

    2014-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2008 of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment-specific laboratory experiments, in-situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site-specific characterization. Because the relationship is important to understanding the evolution of the Yucca Mountain Project, the tabulation also shows the interaction between four broad categories of political bodies and government agencies/institutions: (a) technical milestones of the implementing institutions, (b) development of the regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives and decisions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste.

  2. Method of preparing nuclear wastes for tansportation and interim storage

    Science.gov (United States)

    Bandyopadhyay, Gautam; Galvin, Thomas M.

    1984-01-01

    Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

  3. Third international seabed high-level waste disposal assessment workshop, Albuquerque, New Mexico, February 6--7, 1978: a report to the NEA Radioactive Waste Management Committee

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.R. (ed.)

    1978-10-01

    The task groups of the Third International Workshop were staffed by scientists from the attending countries. Reviews of the progress of programs within each nation were given and plans for cooperative task group workshops, data interchanges, newsletters, ocean cruises, sample exchanges, and critical laboratory and field measurements were coordinated. Although a considerable amount of work remains to be done to assure safety and feasibility, no technical or environmental reasons were identified that would preclude the disposal of radioactive wastes beneeath the ocean floor.

  4. Thermo-Hydro Mechanical Characteristics and Processes in the Clay Barrier of a High Level Radioactive Waste Repository. State of the Art Report

    Energy Technology Data Exchange (ETDEWEB)

    Villar, M. V.

    2004-07-01

    This document is a summary of the available information on the thermo-hydro-mechanical properties of the bentonite barrier of a high-level radioactive waste repository and of the processes taking place in it during the successive repository operation phases. Mainly the thermal properties, the volume change processes (swelling and consolidation), the permeability and the water retention capacity are analysed. A review is made of the existing experimental knowledge on the modification of the these properties by the effect of temperature, water salinity, humidity and density of the bentonite, and their foreseen evolution as a consequence of the processes expected in the repository. The compiled evolution refers mostly to the FEBEX (Spain), the MX-80 (US) and the FoCa (France) bentonite, considered as reference barrier materials in several European disposal concepts. (Author) 102 refs.

  5. Pre-construction geologic section along the cross drift through the potential high-level radioactive waste repository, Yucca Mountain, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    Potter, C.J.; Day, W.C.; Sweetkind, D.S. [Geological Survey, Denver, CO (United States); Juan, C.S.; Drake, R.M. II [Pacific Western Technologies, Ltd., Denver, CO (United States)

    1998-12-31

    As part of the Site Characterization effort for the US Department of Energy`s Yucca Mountain Project, tunnels excavated by tunnel boring machines provide access to the volume of rock that is under consideration for possible underground storage of high-level nuclear waste beneath Yucca Mountain, Nevada. The Exploratory Studies Facility, a 7.8-km-long, 7.6-m-diameter tunnel, has been excavated, and a 2.8-km-long, 5-m-diameter Cross Drift will be excavated in 1998 as part of the geologic, hydrologic and geotechnical evaluation of the potential repository. The southwest-trending Cross Drift branches off of the north ramp of the horseshoe-shaped Exploratory Studies Facility. This report summarizes an interpretive geologic section that was prepared for the Yucca Mountain Project as a tool for use in the design and construction of the Cross Drift.

  6. Geologic and hydrologic characterization and evaluation of the Basin and Range Province relative to the disposal of high-level radioactive waste. Part I. Introduction and guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Bedinger, M.S.; Sargent, K.A.; Reed, J.E.

    1984-12-31

    The US Geological Survey`s program for geologic and hydrologic evaluation of physiographic provinces to identify areas potentially suitable for locating repository sites for disposal of high-level nuclear wastes was announced to the Governors of the eight states in the Basin and Range Province on May 5, 1981. Representatives of Arizona, California, Idaho, New Mexico, Nevada, Oregon, Texas, and Utah, were invited to cooperate with the federal government in the evaluation process. Each governor was requested to nominate an earth scientist to represent the state in a province working group composed of state and US Geological Survey representatives. This report, Part I of a three-part report, provides the background, introduction and scope of the study. This part also includes a discussion of geologic and hydrologic guidelines that will be used in the evaluation process and illustrates geohydrologic environments and the effect of individual factors in providing multiple natural barriers to radionuclide migration. 27 refs., 6 figs., 1 tab.

  7. Geologic and hydrologic characterization and evaluation of the Basin and Range Province relative to the disposal of high-level radioactive waste: Part II, Geologic and hydrologic characterization

    Science.gov (United States)

    Sargent, Kenneth A.; Bedinger, M.S.

    1985-01-01

    The geology and hydrology of the Basin and Range Province of the western conterminous United States are characterized in a series of data sets depicted in maps compiled for evaluation of prospective areas for further study of geohydrologic environments for isolation of high-level radioactive waste. The data sets include: (1) Average precipitation and evaporation; (2) surface distribution of selected rock types; (3) tectonic conditions; and (4) surface- and ground -water hydrology and Pleistocene lakes and marshes.Rocks mapped for consideration as potential host media for the isolation of high-level radioactive waste are widespread and include argillaceous rocks, granitic rocks, tuffaceous rocks, mafic extrusive rocks, evaporites, and laharic breccias. The unsaturated zone, where probably as thick as 150 meters (500 feet), was mapped for consideration as an environment for isolation of high-level waste. Unsaturated rocks of various lithologic types are widespread in the Province.Tectonic stability in the Quaternary Period is considered the key to assessing the probability of future tectonism with regard to high-level radioactive waste disposal. Tectonic conditions are characterized on the basis of the seismic record, heat-flow measurements, the occurrence of Quaternary faults, vertical crustal movement, and volcanic features. Tectonic activity, as indicated by seismicity, is greatest in areas bordering the western margin of the Province in Nevada and southern California, the eastern margin of the Province bordering the Wasatch Mountains in Utah and in parts of the Rio Grande valley. Late Cenozoic volcanic activity is widespread, being greatest bordering the Sierra Nevada in California and Oregon, and bordering the Wasatch Mountains in southern Utah and Idaho.he arid to semiarid climate of the Province results in few perennial streams and lakes. A large part of the surface drainage is interior and the many closed basins commonly are occupied by playas or dry lake

  8. Polysomatism and structural complexity. Structure model for murataite-8C, a complex crystalline matrix for the immobilization of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Pakhomova, Anna S. [Bayreuth Univ. (Germany). Bayerisches Geoinstitut; Krivovichev, Sergey V. [St. Petersburg State Univ. (Russian Federation). Dept. of Crystallography; Yudintsev, Sergey V. [Institute of Geology of Ore deposits, Petrography, Mineralogy and Geochemistry, Moscow (Russian Federation); Stefanovsky, Sergey V. [RAS, Moscow (Russian Federation). A.N. Frumkin Institute of Physical Chemistry and Electrochemistry

    2016-03-15

    Murataite-8C, a prospective synthetic material for the long-term immobilization of high-level radioactive waste and a member of the pyrochlore-murataite polysomatic series, was investigated by means of single-crystal X-ray diffraction. The crystal structure (cubic, F-43m, a = 39.105(12) Aa, V = 59799(32) Aa{sup 3}, Z = 4) is of outstanding complexity and contains forty symmetrically independent cation sites. Its crystal-chemical formula determined on the basis of the crystal-structure refinement, Al{sub 25.44}Ca{sub 55.96}Ti{sub 282.20}Mn{sub 53.72}Fe{sub 17.24}Zr{sub 15.00}Ho{sub 36} {sub .64}O{sub 823}, is in good agreement with the empirical formula calculated from electronmicroprobe data, Al{sub 23.02}Ca{sub 52.85}Ti{sub 284.10}Mn{sub 54.31}Fe{sub 17.59}Zr{sub 14.83}Ho{sub 38} {sub .04}O{sub 823}. The crystal structure is based upon a three-dimensional octahedral framework that can be described as an alternation of murataite and pyrochlore modules immersed into a transitional substructure that combine elements of the crystal structures of murataite-3C and pyrochlore. The obtained structural model confirms the polysomatic nature of the pyrochlore-murataite series and illuminates the chemical and structural peculiarities of crystallization of the murataite-type titanate ceramic matrices. The high chemical and structural complexity of the members of the pyrochlore-murataite series is unparalleled in the world of crystalline materials proposed for the high-level radioactive waste immobilization, which makes it unique and promising for further technological and scientific exploration.

  9. Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Part 3. Safety assessment for geological disposal systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the geological disposal concept for high-level radioactive wastes (HLW) in Japan. The present document, the part 3 of the progress report, concerns safety assessment for geological disposal systems definitely introduced in part 1 and 2 of this series and consists of 9 chapters. Chapter I concerns the methodology for safety assessment while Chapter II deals with diversity and uncertainty about the scenario, the adequate model and the required data of the systems above. Chapter III summarizes the components of the geological disposal system. Chapter IV refers to the relationship between radioactive wastes and human life through groundwater, i.e. nuclide migration. In Chapter V is made a reference case which characterizes the geological environmental data using artificial barrier specifications. (Ohno. S.)

  10. Estimation of the limitations for surficial water addition above a potential high level radioactive waste repository at Yucca Mountain, Nevada; Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Fewell, M.E.; Sobolik, S.R.; Gauthier, J.H.

    1992-01-01

    The Yucca Mountain Site Characterization Project is studying Yucca Mountain in southwestern Nevada as a potential site for a high-level nuclear waste repository. Site characterization includes surface-based and underground testing. Analyses have been performed to design site characterization activities with minimal impact on the ability of the site to isolate waste, and on tests performed as part of the characterization process. One activity of site characterization is the construction of an Exploratory Studies Facility, consisting of underground shafts, drifts, and ramps, and the accompanying surface pad facility and roads. The information in this report addresses the following topics: (1) a discussion of the potential effects of surface construction water on repository-performance, and on surface and underground experiments; (2) one-dimensional numerical calculations predicting the maximum allowable amount of water that may infiltrate the surface of the mountain without affecting repository performance; and (3) two-dimensional numerical calculations of the movement of that amount of surface water and how the water may affect repository performance and experiments. The results contained herein should be used with other site data and scientific/engineering judgement in determining controls on water usage at Yucca Mountain. This document contains information that has been used in preparing Appendix I of the Exploratory Studies Facility Design Requirements document for the Yucca Mountain Site Characterization Project.

  11. Co-metabolic enhancement of organic removal from waste water in the presence of high levels of alkyl paraben constituents of cosmetic and personal care products.

    Science.gov (United States)

    Fan, Chihhao; Wang, Shin-Chih

    2017-07-01

    The enhanced removal of organic material from municipal waste water containing 50 mg/L of chemical oxygen demand and a given amount of alkyl paraben using a biofilm system was investigated. The parabens used were methyl, ethyl, and propyl paraben. The experiments were conducted at influent paraben concentrations of 10 and 50 mg/L. The influent pH was measured around 4.6 because of paraben hydrolysis. The effluent pH increased due to hydrogen consumption and small molecular acid generation. The higher removal rates were observed for the paraben with longer alkyl chains, which were more hydrophobic and capable of penetrating into microbial cells. The co-existing organic constituents in municipal waste water were found to be competitive with paraben molecules for microbial degradation at low paraben loading (i.e., 10 mg/L). Instead, the co-metabolic effect was observed at a higher paraben loading (i.e., 50 mg/L) due to more active enzymatic catalysis, implying the possible enhancement or organic removal in the presence of high levels of parabens. The difference in BOD and TOC removing ratios for parabens decreased with increasing HRT, implying their better mineralization than that of municipal organic constituents. This was because the microbial organism became more adapted to the reacting system with longer HRT, and more oxygenase was produced to facilitate the catechol formation and ring-opening reactions, causing apparent enhancement in mineralization. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Standard practice for prediction of the long-term behavior of materials, including waste forms, used in engineered barrier systems (EBS) for geological disposal of high-level radioactive waste

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade). 1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation. 1.1.2 The predictions are based on models derived from theoretical considerat...

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

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; FENG Z; GAN H; PEGG IL

    2009-11-05

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

  14. Evolution of chemical conditions and estimated solubility controls on radionuclides in the residual waste layer during post-closure aging of high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Denham, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Millings, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2012-08-28

    This document provides information specific to H-Area waste tanks that enables a flow and transport model with limited chemical capabilities to account for varying waste release from the tanks through time. The basis for varying waste release is solubilities of radionuclides that change as pore fluids passing through the waste change in composition. Pore fluid compositions in various stages were generated by simulations of tank grout degradation. The first part of the document describes simulations of the degradation of the reducing grout in post-closure tanks. These simulations assume flow is predominantly through a water saturated porous medium. The infiltrating fluid that reacts with the grout is assumed to be fluid that has passed through the closure cap and into the tank. The results are three stages of degradation referred to as Reduced Region II, Oxidized Region II, and Oxidized Region III. A reaction path model was used so that the transitions between each stage are noted by numbers of pore volumes of infiltrating fluid reacted. The number of pore volumes to each transition can then be converted to time within a flow and transport model. The bottoms of some tanks in H-Area are below the water table requiring a different conceptual model for grout degradation. For these simulations the reacting fluid was assumed to be 10% infiltrate through the closure cap and 90% groundwater. These simulations produce an additional four pore fluid compositions referred to as Conditions A through D and were intended to simulate varying degrees of groundwater influence. The most probable degradation path for the submerged tanks is Condition C to Condition D to Oxidized Region III and eventually to Condition A. Solubilities for Condition A are estimated in the text for use in sensitivity analyses if needed. However, the grout degradation simulations did not include sufficient pore volumes of infiltrating fluid for the grout to evolve to Condition A. Solubility controls for use

  15. One-pot synthesis of carbon supported calcined-Mg/Al layered double hydroxides for antibiotic removal by slow pyrolysis of biomass waste

    Science.gov (United States)

    Tan, Xiaofei; Liu, Shaobo; Liu, Yunguo; Gu, Yanling; Zeng, Guangming; Cai, Xiaoxi; Yan, Zhili; Yang, Chunping; Hu, Xinjiang; Chen, Bo

    2016-12-01

    A biochar supported calcined-Mg/Al layered double hydroxides composite (CLDHs/BC) was synthesized by a one-pot slow pyrolysis of LDHs preloaded bagasse biomass. Multiple characterizations of the product illustrated that the calcined-Mg/Al layered double hydroxides (CLDHs) were successfully coated onto the biochar in slow pyrolysis of pre-treated biomass. The as-synthesized CLDHs/BC could efficiently remove antibiotic tetracycline from aqueous solutions. The coating of CLDHs significantly increased the adsorption ability of biochar, and CLDHs/BC exhibited more than 2 times higher adsorption capacity than that of the pristine biochar (BC) in the tested pH range. The maximum adsorption capacity of CLDHs/BC for tetracycline was 1118.12 mg/g at 318 K. The experimental results suggested that the interaction with LDHs on biochar played a dominant role in tetracycline adsorption, accompanied with π-π interaction and hydrogen bond. This study provides a feasible and simple approach for the preparation of high-performance material for antibiotics contaminated wastewater treatment in a cost-effective way.

  16. CASTOR {sup registered} HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Vossnacke, A.; Klein, K.; Kuehne, B. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany)

    2004-07-01

    Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR {sup registered} HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR {sup registered} HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR {sup registered} HAW28M, meeting these future requirements. The CASTOR {sup registered} HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR {sup registered} HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be

  17. Property/composition relationships for Hanford high-level waste glasses melting at 1150{degrees}C volume 2: Chapters 12-16 and appendices A-K

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation Study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO{sub 2}, B{sub 2}O{sub 3}, ZrO{sub 2}, Na{sub 2}O, Li{sub 2}O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity ({eta}), electrical conductivity ({epsilon}), glass transition temperature (T{sub g}), thermal expansion of solid glass ({alpha}{sub s}) and molten glass ({alpha}{sub m}), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T{sub L}), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r{sub mi}) and the 7-day Product Consistency Test (PCT, r{sub pi}), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T{sub L}) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria.

  18. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    Energy Technology Data Exchange (ETDEWEB)

    P. D. Wheatley (INEEL POC); R. P. Rechard (SNL)

    1998-09-01

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

  19. Property/composition relationships for Hanford high-level waste glasses melting at 115{degrees}C volume 1: Chapters 1-11

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO{sub 2}, B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, Fe{sub 2}O{sub 3}, ZrO{sub 2}, Na{sub 2}O, Li{sub 2}O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity ({eta}), electrical conductivity ({epsilon}), glass transition temperature (T{sub g} ), thermal expansion of solid glass ({alpha}{sub s}) and molten glass ({alpha}{sub m}), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T{sub L}), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r{sub mi}) and the 7-day Product Consistency Test (PCT, r{sub pi}), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T{sub L}) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria.

  20. Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects: Report for April 1986-September 1987

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.E.; Arnold, W.D.; Blencoe, J.G.; O`Kelley, G.D.; Land, J.F.

    1988-02-01

    Experiments were conducted with tuff from the proposed high-level nuclear waste site at Yucca Mountain, Nevada. Batch sorption ratio determinations were conducted for strontium, cesium, uranium, and technetium onto samples of tuff using real and synthetic groundwater J-13. There were no significant differences in sorption ratios in experiments with real and synthetic groundwater. Columns 1 cm in diameter and about 5 cm long were constructed, and experiments were conducted with the objective of correlating the results of batch and the column experiments. The characteristics of the columns were tested by determination of elution curves in J-13 containing tritium and technetium as the TcO{sub 4}{sup -} ion. For strontium and cesium, fairly good correlation between values of the sorption ratio obtained by the two methods was observed. Little or no technetium sorption was observed with either method. The elution peaks obtained with neptunium and uranium were asymmetrical and the shapes were often complex, observations which suggest irreversibilities in the sorption reaction. An experiment was performed to provide information on the compositions of the first groundwaters that will contact waste canisters in a tuff-hosted repository after very near field temperatures have cooled to below 100{degree}C. Synthetic groundwater J-13 was slowly dripped onto a slab of tuff maintained at 95-100{degree}C, and the result was a thin encrustation of solids on the slab as the water evaporated. Fresh J-13 groundwater was then allowed to contact the encrustation in a vessel maintained at 90{degree}C. The principal result of the experiment was a significant loss of calcium and magnesium from the fresh J-13 groundwater.

  1. Geologic and hydrologic characterization and evaluation of the Basin and Range Province relative to the disposal of high-level radioactive waste: Part I, Introduction and guidelines

    Science.gov (United States)

    Bedinger, M.S.; Sargent, Kenneth A.; Reed, J.E.

    1984-01-01

    The U.S. Geological Survey's program for geologic and hydrologic evaluation of physiographic provinces to identify areas potentially suitable for locating repository sites for disposal of high-level nuclear wastes was announced to the Governors of the eight States in the Basin and Range Province on May 5, 1981. Representatives of Arizona, California, Idaho, New Mexico, Nevada, Oregon, Texas, and Utah, were invited to cooperate with the Federal Government in the evaluation process. Each Governor was requested to nominate an Earth scientist to represent the State in a province working group composed of State and U.S. Geological Survey representatives. This report, Part I of a three-part report, provides the background, introduction and scope of the study. This part also includes a discussion of geologic and hydrologic guidelines that will be used in the evaluation process and illustrates geohydrologic environments and the effect of individual factors in providing multiple natural barriers to radionuclide migration.Part II is a reconnaissance characterization of the geologic and hydrologic factors to be used in the initial screening of the Basin and Range Province. Part III will be the initial evaluation of the Province and will identify regions that appear suitable for further study.The plan for study of the Province includes a stepwise screening process by which successively smaller land units are considered in increasing detail. Each step involves characterization of the geology and hydrology and selection of subunits for more intensive characterization. Selection of subunits for further study is by evaluation of geologic and hydrologic conditions following a set of guidelines. By representation on the Province Working Group, the States participate in a consultation and review role in: (1) Establishing geologic and hydrologic guidelines, and (2) characterizing and evaluating the Province. The States also participate in compilation of geologic and hydrologic data

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hua, E-mail: nzhangh@aliyun.com [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); China Institute of Atomic Energy, P.O. Box 275-93, 102413 Beijing (China); Corkhill, Claire L.; Heath, Paul G.; Hand, Russell J.; Stennett, Martin C. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); Hyatt, Neil C., E-mail: n.c.hyatt@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2015-07-15

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

  3. Illustration of sampling-based approaches to the calculation of expected dose in performance assessments for the proposed high level radioactive waste repository at Yucca Mountain, Nevada.

    Energy Technology Data Exchange (ETDEWEB)

    Helton, Jon Craig (Arizona State University, Tempe, AZ); Sallaberry, Cedric J. PhD. (.; .)

    2007-04-01

    A deep geologic repository for high level radioactive waste is under development by the U.S. Department of Energy at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the U.S. Environmental Protection Agency (EPA) has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the U.S. Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This presentation describes and illustrates how general and typically nonquantitive statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are poorly known but assumed to have constant values in the calculation of expected dose to the RMEI).

  4. Effects of Fuel to Synthesis of CaTiO3 by Solution Combustion Synthesis for High-Level Nuclear Waste Ceramics.

    Science.gov (United States)

    Jung, Choong-Hwan; Kim, Yeon-Ku; Han, Young-Min; Lee, Sang-Jin

    2016-02-01

    A solution combustion process for the synthesis of perovskite (CaTiO3) powders is described. Perovskite is one of the crystalline host matrics for the disposal of high-level radioactive wastes (HLW) because it immobilizes Sr and Lns elements by forming solid solutions. Solution combustion synthesis, which is a self-sustaining oxi-reduction reaction between nitrate and organic fuel, the exothermic reaction, and the heat evolved convert the precursors into their corresponding oxide products above 1100 degrees C in air. To investigate the effects of amino acid on the combustion reaction, various types of fuels were used; a glycine, amine and carboxylic ligand mixture. Sr, La and Gd-nitrate with equivalent amounts of up to 20% of CaTiO3 were mixed with Ca and Ti nitrate and amino acid. X-ray diffraction analysis, SEM and TEM were conducted to confirm the formed phases and morphologies. While powders with an uncontrolled shape are obtained through a general oxide-route process, Ca(Sr, Lns)TiO3 powders with micro-sized soft agglomerates consisting of nano-sized primary particles can be prepared using this method.

  5. Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Part 2. Engineering technology for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the deep geological disposal concept for high-level radioactive wastes (HLW) in Japan. The present document, part 2 of the progress report, concerns engineering aspect with reference to Japanese geological disposal plan, according to which the vitrified HLW will be disposed of into a deep, stable rock mass with thick containers and surrounding buffer materials at the depth of several hundred meters. It discusses on multi-barrier systems consisting of a series of engineered and natural barriers that will isolate radioactive nuclides effectively and retard their migrations to the biosphere environment. Performance of repository components, including specifications of containers for vitrified HLW and their overpacks under design as well as buffer material such as Japanese bentonite to be placed in between are described referring also to such possible problems as corrosion arising from the supposed system. It also presents plans and designs for underground disposal facilities, and the presumed management of the underground facilities. (Ohno, S.)

  6. A literature review of coupled thermal-hydrologic-mechanical-chemical processes pertinent to the proposed high-level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Manteufel, R.D.; Ahola, M.P.; Turner, D.R.; Chowdhury, A.H. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1993-07-01

    A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mechanisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or performance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be important, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic reposit``. Although this work stems from regulatory interest in the design of the geologic repository, it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge) to determine which couplings are important, and identify which computer codes are currently available to model coupled processes.

  7. Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Introductory part and summaries

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the geological disposal concept for high-level radioactive wastes (HLW) in Japan and comprises seven chapters. Chapter I briefly describes the importance of HLW management in promoting nuclear energy utilization. According to the long-term program, the HLW separated from spent fuels at reprocessing plants is to be vitrified and stored for a period of 30 to 50 years to allow cooling, then be disposed of in a deep geological formation. Chapter II mainly explains the concepts of geological disposal in Japan. Chapters III to V are devoted to discussions on three important technical elements (the geological environment of Japan, engineering technology and safety assessment of the geological disposal system) which are necessary for reliable realization of the geological disposal concept. Chapter VI demonstrates the technical ground for site selection and for setup of safety standards of the disposal. Chapter VII summarizes together with plans for future research and development. (Ohno, S.)

  8. Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects: Report for April 1986--September 1987

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.E.; Arnold, W.D.; Blencoe, J.G.; O`Kelley, G.D.; Land, J.F.

    1988-07-01

    During this report period, all experiments were conducted with tuff from the proposed high-level nuclear waste site at Yucca Mountain, Nevada. Batch sorption ratio determinations were conducted for strontium, cesium, uranium, and technetium onto samples of tuff using real and synthetic groundwater J-13. There were no significant differences in sorption ratios in experiments with real and synthetic groundwater. Columns were tested by determination of elution curves in J-13 containing tritium and technetium as the TcO{sub 4}/sup {minus}/ ion. For strontium and cesium, fairly good correlation between values of the sorption ratio obtained by the two methods was observed. Little technetium sorption was observed with either method. The elution peaks obtained with neptunium and uranium were asymmetrical and the shapes were often complex, observations which suggest irreversibilities in the sorption reaction. Synthetic groundwater J-13 was slowly dripped onto a slab of tuff maintained at 95--100{degree}C, and the result was a thin encrustation of solids on the slab as the water evaporated. Fresh J-13 groundwater was then allowed to contact the encrustation in a vessel maintained at 90{degree}C. The principal result of the experiment was a significant loss of calcium and magnesium from the fresh J-13 groundwater. 13 refs. 25 figs., 9 tabs.

  9. Development of an Insight Model for Post-closure Radiological Risks from the Disposal of High-level Waste and Spent Fuel

    Science.gov (United States)

    Kelly, M.; Carter, A.; Bailey, L.; Wellstead, M.

    2012-04-01

    Geological disposal is the UK policy for the long-term management of higher activity radioactive wastes. The Radioactive Waste Management Directorate (RWMD) of the Nuclear Decommissioning Authority (NDA) has been given the responsibility for implementing geological disposal in the UK. The approach adopted for building confidence in long-term safety of the GDF is to use multiple barriers to isolate and contain the wastes. NDA-RWMD proposes to develop a safety case based on varied and different lines of reasoning, including both quantitative aspects and qualitative safety arguments. This paper describes the development of a simplified model for high-level waste (HLW) and spent fuel (SF). Assessments of HLW and SF disposals need to reflect a number of physical and chemical processes that govern the release and transport of the disposed wastes. Physical processes include container failure, release of radionuclides from the wasteform, diffusion through bentonite barriers and transport in fractured rock. Chemical processes include solubility limitation and sorption. HLW and SF assessments are generally based on numerical models of these processes, implemented in software tools such as GoldSim. The results of such assessments can be complex to understand, especially in situations where no single release or transport process is dominant. A simplified model of the release and transport of radionuclides for disposals of HLW and SF has been developed. The objectives of the model are to identify the key physical and chemical processes that govern radiological risk, and to enable numerical estimates of risk (and other intermediate quantities) to be made. The model is based on the notion that the properties of the time-dependent model outputs can be characterised in terms of "moments". The moments are easy to compute since they are related in a simple way to the Laplace transforms of the functions under consideration. The Laplace transform method is a standard technique for

  10. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2009 of the performance of a repository for spent nuclear fuel and high - level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment - specific laboratory experiments, in - situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site - specific characterization . The current sixth period beyond 2010 represents a new effort to set waste management policy in the United States. Because the relationship is important to understanding the evolution of the Yucca Mountain Project , the tabulation also shows the interaction between the policy realm and technical realm using four broad categories of events : (a) Regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives, (c) technical milestones of implementing institutions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste. Preface The historical progression of technical milestones for the Yucca Mountain Project was originally developed for 10 journal articles in a special issue of Reliability Engineering System Safety on the performance assessment for the Yucca Mountain license

  11. Potential dispositioning flowsheets for ICPP SNF and wastes

    Energy Technology Data Exchange (ETDEWEB)

    Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

  12. Near Field sorption Data Bases for Compacted MX-80 Bentonite for Performance Assessment of a High-Level Radioactive Waste Repository in Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.; Baeyens, B

    2003-08-01

    Bentonites of various types and compacted forms are being investigated in many countries as backfill materials in high-level radioactive waste disposal concepts. Nagra is currently considering an Opalinus clay (OPA) formation in the Zuercher Weinland as a potential location for a high-level radioactive waste repository. A compacted MX-80 bentonite is foreseen as a potential backfill material. Performance assessment studies will be performed for this site and one of the requirements for such an assessment are sorption data bases (SDB) for the bentonite near-field. The purpose of this report is to describe the procedures used to develop the SDB. One of the pre-requisites for developing a SDB is a water chemistry for the compacted bentonite porewater. For a number of reasons mentioned in the report, and discussed in more detail elsewhere, this is not a straightforward task. There are considerable uncertainties associated with the major ion concentrations and in particular with the system pH and Eh. The MX-80 SDB was developed for a reference bentonite porewater (pH = 7.25) which was calculated using the reference OPA porewater. In addition, two further SDBs are presented for porewaters calculated at pH values of 6.9 and 7.9 corresponding to lower and upper bound values calculated for the range of groundwater compositions anticipated for the OPA host rock. 'In house' sorption isotherm data were measured for Cs(I), Ni(II), Eu(III), Th(IV), Se(IV) and 1(-1) on the 'as received' MX-80 material equilibrated with a simulated porewater composition. Complementary 'in house' sorption edge and isotherm measurements on conditioned Na/Ca montmorillonites were also available for many of these radionuclides. These data formed the core of the SDB. Nevertheless, some of the required sorption data still had to be obtained from the open literature. An important part of this report is concerned with describing selection procedures and the modifications

  13. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.

  14. Radiation-Induced Defects in Kaolinite as Tracers of Past Occurrence of Radionuclides in a Natural Analogue of High Level Nuclear Waste Repository

    Science.gov (United States)

    Allard, T.; Fourdrin, C.; Calas, G.

    2007-05-01

    Understanding the processes controlling migrations of radioelements at the Earth's surface is an important issue for the long-term safety assessment of high level nuclear waste repositories (HLNWR). Evidence of past occurrence and transfer of radionuclides can be found using radiation-induced defects in minerals. Clay minerals are particularly relevant because of their widespread occurrence at the Earth's surface and their finely divided nature which provides high contact area with radioactive fluids. Owing to its sensitivity to radiations, kaolinite can be used as natural, in situ dosimeter. Kaolinite is known to contain radiation-induced defects which are detected by Electron Paramagnetic Resonance. They are differentiated by their nature, their production kinetics and their thermal stability. One of these defects is stable at the scale of geological periods and provides a record of past radionuclide occurrence. Based on artificial irradiations, a methodology has been subsequently proposed to determine paleodose cumulated by kaolinite since its formation. The paleodose can be used to derive equivalent radioelement concentrations, provided that the age of kaolinite formation can be constrained. This allows quantitative reconstruction of past transfers of radioelements in natural systems. An example is given for the Nopal I U-deposit (Chihuahua, Mexico), hosted in hydrothermally altered volcanic tufs and considered as analogue of the Yucca Mountain site. The paleodoses experienced by kaolinites were determined from the concentration of defects and dosimetry parameters of experimental irradiations. Using few geochemical assumption, a equivalent U-content responsible for defects in kaolinite was calculated from the paleodose, a dose rate balance and model ages of kaolinites constrained by tectonic phases. In a former study, the ages were assumptions derived from regional tectonic events. In thepresent study, ages of mineralization events are measured from U

  15. Final Systems Development Report for the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mountain, NV

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-06-18

    The Systems Development Report represents the third major step in the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mound Nevada. The first of these steps was to forge a Research Design that would serve as a guide for the overall research process. The second step was the construction of the Base Case, the purpose of which was to describe existing conditions in Clark County in the specified analytic areas of Economic-Demographic/Fiscal, Emergency Planning and Management, Transportation and Sociocultural analysis. The base case description will serve as a basis for assessing changes in these topic areas that might result from the Yucca Mountain project. These changes will be assessed by analyzing conditions with and without repository development in the county. Prior to performing such assessments, however, the snapshot type of data found in the base case must be operationalized or systematized to allow for more dynamic data utilization. In other words, a data system that can be used to analyze the consequences of the introduction of different variables (or variable values) in the Clark County context must be constructed. Such a system must be capable of being updated through subsequent data collection and monitoring efforts to both provide a rolling base case and supply information necessary to construct trend analyses. For example, during the Impact Assessment phase of the study process, the without repository analysis is accomplished by analyzing growth for the county given existing conditions and likely trends. These data are then compared to the with Yucca Mountain project conditions anticipated for the county. Similarly, once the emergency planning management and response needs associated with the repository are described, these needs will be juxtaposed against existing (and various future) capacity(ies) in order to determine the nature and magnitude of impacts in this analytic area. Analogous tasks

  16. DOCUMENTATION OF NATIONAL WEATHER CONDITIONS AFFECTING LONG-TERM DEGRADATION OF COMMERCIAL SPENT NUCLEAR FUEL AND DOE SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE

    Energy Technology Data Exchange (ETDEWEB)

    W. L. Poe, Jr.; P.F. Wise

    1998-11-01

    The U.S. Department of Energy (DOE) is preparing a proposal to construct, operate 2nd monitor, and eventually close a repository at Yucca Mountain in Nye County, Nevada, for the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). As part of this effort, DOE has prepared a viability assessment and an assessment of potential consequences that may exist if the repository is not constructed. The assessment of potential consequences if the repository is not constructed assumes that all SNF and HLW would be left at the generator sites. These include 72 commercial generator sites (three commercial facility pairs--Salem and Hope Creek, Fitzpatrick and Nine Mile Point, and Dresden and Morris--would share common storage due to their close proximity to each other) and five DOE sites across the country. DOE analyzed the environmental consequences of the effects of the continued storage of these materials at these sites in a report titled Continued Storage Analysis Report (CSAR; Reference 1 ) . The CSAR analysis includes a discussion of the degradation of these materials when exposed to the environment. This document describes the environmental parameters that influence the degradation analyzed in the CSAR. These include temperature, relative humidity, precipitation chemistry (pH and chemical composition), annual precipitation rates, annual number of rain-days, and annual freeze/thaw cycles. The document also tabulates weather conditions for each storage site, evaluates the degradation of concrete storage modules and vaults in different regions of the country, and provides a thermal analysis of commercial SNF in storage.

  17. Geology of the Yucca Mountain Region, Chapter in Stuckless, J.S., ED., Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    J.S. Stuckless; D. O' Leary

    2006-09-25

    Yucca Mountain has been proposed as the site for the Nation's first geologic repository for high-level radioactive waste. This chapter provides the geologic framework for the Yucca Mountain region. The regional geologic units range in age from late Precambrian through Holocene, and these are described briefly. Yucca Mountain is composed dominantly of pyroclastic units that range in age from 11.4 to 15.2 Ma. The proposed repository would be constructed within the Topopah Spring Tuff, which is the lower of two major zoned and welded ash-flow tuffs within the Paintbrush Group. The two welded tuffs are separated by the partly to nonwelded Pah Canyon Tuff and Yucca Mountain Tuff, which together figure prominently in the hydrology of the unsaturated zone. The Quaternary deposits are primarily alluvial sediments with minor basaltic cinder cones and flows. Both have been studied extensively because of their importance in predicting the long-term performance of the proposed repository. Basaltic volcanism began about 10 Ma and continued as recently as about 80 ka with the eruption of cones and flows at Lathrop Wells, approximately 10 km south-southwest of Yucca Mountain. Geologic structure in the Yucca Mountain region is complex. During the latest Paleozoic and Mesozoic, strong compressional forces caused tight folding and thrust faulting. The present regional setting is one of extension, and normal faulting has been active from the Miocene through to the present. There are three major local tectonic domains: (1) Basin and Range, (2) Walker Lane, and (3) Inyo-Mono. Each domain has an effect on the stability of Yucca Mountain.

  18. Final base case community analysis: Indian Springs, Nevada for the Clark County socioeconomic impact assessment of the proposed high- level nuclear waste repository at Yucca Mountain, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-06-18

    This document provides a base case description of the rural Clark County community of Indian Springs in anticipation of change associated with the proposed high-level nuclear waste repository at Yucca Mountain. As the community closest to the proposed site, Indian Springs may be seen by site characterization workers, as well as workers associated with later repository phases, as a logical place to live. This report develops and updates information relating to a broad spectrum of socioeconomic variables, thereby providing a `snapshot` or `base case` look at Indian Springs in early 1992. With this as a background, future repository-related developments may be analytically separated from changes brought about by other factors, thus allowing for the assessment of the magnitude of local changes associated with the proposed repository. Given the size of the community, changes that may be considered small in an absolute sense may have relatively large impacts at the local level. Indian Springs is, in many respects, a unique community and a community of contrasts. An unincorporated town, it is a small yet important enclave of workers on large federal projects and home to employees of small- scale businesses and services. It is a rural community, but it is also close to the urbanized Las Vega Valley. It is a desert community, but has good water resources. It is on flat terrain, but it is located within 20 miles of the tallest mountains in Nevada. It is a town in which various interest groups diverge on issues of local importance, but in a sense of community remains an important feature of life. Finally, it has a sociodemographic history of both surface transience and underlying stability. If local land becomes available, Indian Springs has some room for growth but must first consider the historical effects of growth on the town and its desired direction for the future.

  19. Investigation on Calcination Behaviors of Coal Gangue by Fluidized Calcination in Comparison with Static Calcination

    Directory of Open Access Journals (Sweden)

    Shuai Yuan

    2017-02-01

    Full Text Available In order to comprehensively utilize coal gangue, we present fluidized calcination as a new thermal technology for activating coal gangue and systematical study was conducted in comparison with static calcination. The calcined products obtained by different calcination methods under various temperatures were characterized by the means of X-ray diffraction (XRD, thermal gravimetry-differential scanning calorimeter (TG-DSC, Fourier transform-infrared spectroscopy (FT-IR and scanning electron microscope-energy dispersive spectrometer (SEM-EDS. Chemical and physical characteristics such as aluminium leaching rate, chemical oxygen demand and whiteness of calcined products were also investigated. The results show that aluminium leaching rate could reach to the maximal value 74.42% at 500 °C by fluidized calcination, while the maximal value of 66.33% could be reached at 600 °C by static calcination. Products by fluidized calcination obtained higher whiteness and lower chemical oxygen demand (COD under the same calcination temperature. The well-crystallized kaolinite transform to amorphous meta-kaolinite under 600 °C and mullite presence under 1000 °C according to phase transformation, chemical bond variation and microstructure evolution analysis. Fluidized calcination was more efficiently for combustion of carbon/organic matter and dehydroxylation of kaolinite, which might applied in coal gangue industry in future.

  20. Evaluation of coupled thermo-hydro-mechanical phenomena in the near field for geological disposal of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Chijimatsu, Masakazu; Fujita, Tomoo; Sugita, Yutaka; Taniguchi, Wataru [Japan Nuclear Cycle Development Inst., Tokai Works, Waste Management and Fuel Cycle Research Center, Waste Isolation Research Division, Barrier Performance Group, Tokai, Ibaraki (Japan)

    2000-01-01

    Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraulic conductivity and high adsorption capacity of radionuclides. In a repository of HLW, complex thermal, hydraulic and mechanical (T-H-M) phenomena will take place, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of ground water and stress generation due to the earth pressure, the thermal loading and the swelling pressure of the buffer material. In order to evaluate the performance of the buffer material, the coupled T-H-M behaviors within the compacted bentonite have to be modelled. Before establishing a fully coupled T-H-M model, the mechanism of each single phenomenon or partially coupled phenomena should be identified. Furthermore, in order to evaluate the coupled T-H-M phenomena, the analysis model was developed physically and numerically and the adequacy and the applicability was tested though the engineered scale laboratory test and in-situ test. In this report, the investigative results for the development of coupled T-H-M model were described. This report consists of eight chapters. In Chapter 1, the necessity of coupled T-H-M model in the geological disposal project of the high-level radioactive waste was described . In Chapter 2, the laboratory test results of the rock sample and the buffer material for the coupled T-H-M analysis were shown. The rock samples were obtained from the in-situ experimental site at Kamaishi mine. As the buffer material, bentonite clay (Kunigel V1 and Kunigel OT-9607) and bentonite-sand mixture were used. In Chapter 3, in-situ tests to obtain the rock property were shown. As in-situ tests

  1. Studies on Am(III) separation from simulated high-level waste using cobalt bis(dicarbollide) (1(-)) ion derivative covalently bound to N,N'-di-n-octyl diglycol diamide as extractant and DTPA as stripping agent

    Czech Academy of Sciences Publication Activity Database

    Bubeníková, M.; Selucký, P.; Rais, J.; Grüner, Bohumír; Švec, Petr

    2012-01-01

    Roč. 293, č. 1 (2012), s. 403-408 ISSN 0236-5731 R&D Projects: GA ČR GA104/09/0668 Institutional research plan: CEZ:AV0Z40320502 Keywords : Solvent extraction * actinides * high-level liquid waste * dicarbollide derivatives * carboranes * TODGA * DTPA Subject RIV: CA - Inorganic Chemistry Impact factor: 1.467, year: 2012

  2. High Temperature Calcination - MACT Upgrade Equipment Pilot Plant Test

    Energy Technology Data Exchange (ETDEWEB)

    Richard D. Boardman; B. H. O& #39; Brien; N. R. Soelberg; S. O. Bates; R. A. Wood; C. St. Michel

    2004-02-01

    About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste are stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Calcination at high-temperature conditions (600 C, with alumina nitrate and calcium nitrate chemical addition to the feed) is one of four options currently being considered by the Department of Energy for treatment of the remaining tank wastes. If calcination is selected for future processing of the sodium-bearing waste, it will be necessary to install new off-gas control equipment in the New Waste Calcining Facility (NWCF) to comply with the Maximum Achievable Control Technology (MACT) standards for hazardous waste combustors and incinerators. This will require, as a minimum, installing a carbon bed to reduce mercury emissions from their current level of up to 7,500 to <45 {micro}g/dscm, and a staged combustor to reduce unburned kerosene fuel in the off-gas discharge to <100 ppm CO and <10 ppm hydrocarbons. The staged combustor will also reduce NOx concentrations of about 35,000 ppm by 90-95%. A pilot-plant calcination test was completed in a newly constructed 15-cm diameter calciner vessel. The pilot-plant facility was equipped with a prototype MACT off-gas control system, including a highly efficient cyclone separator and off-gas quench/venturi scrubber for particulate removal, a staged combustor for unburned hydrocarbon and NOx destruction, and a packed activated carbon bed for mercury removal and residual chloride capture. Pilot-plant testing was performed during a 50-hour system operability test January 14-16, followed by a 100-hour high-temperature calcination pilot-plant calcination run January 19-23. Two flowsheet blends were tested: a 50-hour test with an aluminum-to-alkali metal molar ratio (AAR) of 2.25, and a 50-hour test with an AAR of 1.75. Results of the testing

  3. Proposed methodology for completion of scenario analysis for the Basalt Waste Isolation Project. [Assessment of post-closure performance for a proposed repository for high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Roberds, W.J.; Plum, R.J.; Visca, P.J.

    1984-11-01

    This report presents the methodology to complete an assessment of postclosure performance, considering all credible scenarios, including the nominal case, for a proposed repository for high-level nuclear waste at the Hanford Site, Washington State. The methodology consists of defensible techniques for identifying and screening scenarios, and for then assessing the risks associated with each. The results of the scenario analysis are used to comprehensively determine system performance and/or risk for evaluation of compliance with postclosure performance criteria (10 CFR 60 and 40 CFR 191). In addition to describing the proposed methodology, this report reviews available methodologies for scenario analysis, discusses pertinent performance assessment and uncertainty concepts, advises how to implement the methodology (including the organizational requirements and a description of tasks) and recommends how to use the methodology in guiding future site characterization, analysis, and engineered subsystem design work. 36 refs., 24 figs., 1 tab.

  4. Far Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradburry, M.; Baeyens, B

    2003-08-01

    An Opalinus Clay formation in the Zuercher Weinland is under consideration by Nagra as a potential location for a high-level and long-Iived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations. In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/P{sub CO{sub 2}}. In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases. The basis for the sorption data bases is 'in-house' sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zuercher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation . Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (Rd) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature Rd values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting Rd values were then further modified using so-called Lab{yields}Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions. Finally, it is important to have some

  5. Studies on the feasibility of using completely incinerable reagents for the single-cycle separation of americium(III) from simulated high-level liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, P.K.; Kumaresan, R.; Venkatesan, K.A.; Subramanian, G.G.S.; Prathibha, T.; Syamala, K.V.; Selvan, B. Robert; Rajeswari, S.; Antony, M.P.; Rao, P.R. Vasudeva [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.; Chaurasia, Shivkumar; Bhanage, B.M. [Institute of Chemical Technology, Mumbai (India)

    2015-06-01

    The extraction and stripping behavior of various metal ions present in the fast reactor simulated high-level liquid waste (FR-SHLLW) was studied using a solvent phase composed of a neutral extractant, N,N,-didodecyl-N',N'-dioctyl-3-oxapentane-1,5-diamide (D{sup 3}DODGA) and an acidic extractant, di-2-ethylhexyl diglycolamic acid (HDEHDGA) in n-dodecane (n-DD). The third phase formation behavior of the solvent formulation D{sup 3}DODGA + HDEHDGA/n-DD, was studied when it was contacted with FR-SHLLW, and the concentration of neutral and acidic extractant needed to avoid the third phase formation was optimized. The distribution ratio of various metal ions present in FR-SHLLW was measured in a solution of 0.1 M D{sup 3}DODGA + 0.2 M HDEHDGA/n-DD. The extraction of Am(III) was accompanied by the co-extraction of lanthanides and unwanted metal ions such as Zr(IV), Y(III), and Pd(II). A procedure was developed to minimize the extraction of unwanted metal ions by using aqueous soluble complexing agents in FR-SHLLW. Based on those results, the counter-current mixer-settler run was performed in a 20-stage mixer-settler. Quantitative extraction of Am(III), Ln(III), Y(III), and Sr(II) in 0.1 M D{sup 3}DODGA + 0.2 M HDEHDGA/n-DD was observed. The recovery of Am(III) from the loaded organic phase was carried out by the optimized aqueous formulation composed of 0.01 M diethylenetriaminepentaacetic acid (DTPA) + 0.5 M citric acid (CA) at pH 1.5. The stripping of Am(III) was accompanied by co-stripping of some early lanthanides. However the later lanthanides (Eu(III) and beyond) were not back extracted to Am(III) product. Therefore, the studies foresee the possibility of intra-lanthanides as well as lanthanide-actinide separation in a single-processing cycle.

  6. Vitrified waste option study report

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, D.A.; Kimmitt, R.R.

    1998-02-01

    A {open_quotes}Settlement Agreement{close_quotes} between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032.

  7. Conceptual Design of a Simplified Skid-Mounted Caustic-Side Solvent Extraction Process for Removal of Cesium from Savannah Rive Site High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Birdwell, JR.J.F.

    2004-05-12

    This report presents the results of a conceptual design of a solvent extraction process for the selective removal of {sup 137}Cs from high-level radioactive waste currently stored in underground tanks at the U.S. Department of Energy's Savannah River Site (SRS). This study establishes the need for and feasibility of deploying a simplified version of the Caustic-Side Solvent Extraction (CSSX) process; cost/benefit ratios ranging from 33 to 55 strongly support the considered deployment. Based on projected compositions, 18 million gallons of dissolved salt cake waste has been identified as having {sup 137}Cs concentrations that are substantially lower than the worst-case design basis for the CSSX system that is to be deployed as part of the Salt Waste Processing Facility (SWPF) but that does not meet the waste acceptance criteria for immobilization as grout in the Saltstone Manufacturing and Disposal Facility at SRS. Absent deployment of an alternative cesium removal process, this material will require treatment in the SWPF CSSX system, even though the cesium decontamination factor required is far less than that provided by that system. A conceptual design of a CSSX processing system designed for rapid deployment and having reduced cesium decontamination factor capability has been performed. The proposed accelerated-deployment CSSX system (CSSX-A) has been designed to have a processing rate of 3 million gallons per year, assuming 90% availability. At a more conservative availability of 75% (reflecting the novelty of the process), the annual processing capacity is 2.5 million gallons. The primary component of the process is a 20-stage cascade of centrifugal solvent extraction contactors. The decontamination and concentration factors are 40 and 15, respectively. The solvent, scrub, strip, and wash solutions are to have the same compositions as those planned for the SWPF CSSX system. As in the SWPF CSSX system, the solvent and scrub flow rates are equal. The system

  8. High-level waste inventory, characteristics, generation, and facility assessment for treatment, storage, and disposal alternatives considered in the US Department of Energy eenvironmental management programmatic environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Folga, S.M.; Conzelmann, G.; Gillette, J.L.; Kier, P.H.; Poch, L.A.

    1996-12-01

    This report provides data and information needed to support the risk and impact assessments of high-level waste (HLW) management alternatives in the U.S. Department of Energy Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). Available data on the physical form, chemical and isotopic composition, storage locations, and other waste characteristics of interest are presented. High-level waste management follows six implementation phases: current storage, retrieval, pretreatment, treatment, interim canister storage, and geologic repository disposal; pretreatment, treatment, and repository disposal are outside the scope of the WM PEIS. Brief descriptions of current and planned HLW management facilities are provided, including information on the type of waste managed in the facility, costs, product form, resource requirements, emissions, and current and future status. Data sources and technical and regulatory assumptions are identified. The range of HLW management alternatives (including decentralized, regionalized, and centralized approaches) is described. The required waste management facilities include expanded interim storage facilities under the various alternatives. Resource requirements for construction (e.g., land and materials) and operation (e.g., energy and process chemicals), work force, costs, effluents, design capacities, and emissions are presented for each alternative.

  9. Work plan for the identification of techniques for in-situ sensing of layering/interfaces of Hanford high level waste tank

    Energy Technology Data Exchange (ETDEWEB)

    Vargo, G.F. Jr.

    1995-06-16

    The purpose of this work scope is to identify a specific potential technology/device/instrument/ideas that would provide the tank waste data. A method is needed for identifying layering and physical state within the large waste tanks at the Hanford site in Washington State. These interfaces and state changes can adversely impact sampling and characterization activities.

  10. A comparison between ultraviolet disinfection and copper alginate beads within a vortex bioreactor for the deactivation of bacteria in simulated waste streams with high levels of colour, humic acid and suspended solids.

    Directory of Open Access Journals (Sweden)

    Simon F Thomas

    Full Text Available We show in this study that the combination of a swirl flow reactor and an antimicrobial agent (in this case copper alginate beads is a promising technique for the remediation of contaminated water in waste streams recalcitrant to UV-C treatment. This is demonstrated by comparing the viability of both common and UV-C resistant organisms in operating conditions where UV-C proves ineffective - notably high levels of solids and compounds which deflect UV-C. The swirl flow reactor is easy to construct from commonly available plumbing parts and may prove a versatile and powerful tool in waste water treatment in developing countries.

  11. Crystallographic transformation of limestone during calcination under CO2.

    Science.gov (United States)

    Valverde, Jose Manuel; Medina, Santiago

    2015-09-14

    The calcination reaction of limestone (CaCO3) to yield lime (CaO) is at the heart of many industrial applications as well as natural processes. In the recently emerged calcium-looping technology, CO2 capture is accomplished by the carbonation of CaO in a gas-solid reactor (carbonator). CaO is derived by the calcination of limestone in a calciner reactor under necessarily high CO2 partial pressure and high temperature. In situ X-ray diffraction (XRD) has been employed in this work to gain further insight into the crystallographic transformation that takes place during the calcination of limestone under CO2, at partial pressures (P) close to the equilibrium pressure (Peq) and at high temperature. Calcination under these conditions becomes extremely slow. The in situ XRD analysis presented here suggests the presence of an intermediate metastable CaO* phase stemming from the parent CaCO3 structure. According to the reaction mechanism proposed elsewhere, the exothermicity of the CaO* → CaO transformation and high values of P/Peq inhibit the nucleation of CaO at high temperatures. The wt% of CaO* remains at a relatively high level during slow calcination. Two diverse stages have been identified in the evolution of CaO crystallite size, L. Initially, L increases with CaCO3 conversion, following a logarithmic law. Slow calcination allows the crystallite size to grow up from a few nanometers at nucleation up to around 100 nm near the end of conversion. Otherwise, quick calcination at relatively lower CO2 concentrations limits CaO crystallite growth. Once calcination reaches an advanced state, the presence of CaO* drops to zero and the rate of increase of the CaO crystallite size is significantly hindered. Arguably, the first stage in CaO crystallite growth is driven by aggregation of the metastable CaO* nanocrystals, due to surface attractive forces, whereas the second one is consistent with sintering of the aggregated CaO crystals, and persists with time after full

  12. An Assessment of the Stability and the Potential for In-Situ Synthesis of Regulated Organic Compounds in High Level Radioactive Waste Stored at Hanford, Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Wiemers, K.D.; Babad, H.; Hallen, R.T.; Jackson, L.P.; Lerchen, M.E.

    1999-01-04

    The stability assessment examined 269 non-detected regulated compounds, first seeking literature references of the stability of the compounds, then evaluating each compound based upon the presence of functional groups using professional judgment. Compounds that could potentially survive for significant periods in the tanks (>1 year) were designated as stable. Most of the functional groups associated with the regulated organic compounds were considered unstable under tank waste conditions. The general exceptions with respect to functional group stability are some simple substituted aromatic and polycyclic aromatic compounds that resist oxidation and the multiple substituted aliphatic and aromatic halides that hydrolyze or dehydrohalogenate slowly under tank waste conditions. One-hundred and eighty-one (181) regulated, organic compounds were determined as likely unstable in the tank waste environment.

  13. Studies of geology and hydrology in the Basin and Range Province, Southwestern United States, for isolation of high-level radioactive waste - Basis of characterization and evaluation

    Science.gov (United States)

    Bedinger, M.S.; Sargent, K.A.; Langer, William H.; Sherman, Frank B.; Reed, J.E.; Brady, B.T.

    1989-01-01

    The geologic and hydrologic factors in selected regions of the Basin and Range province were examined to identify prospective areas for further study that may provide isolation of high-level radioactive waste from the accessible environment. The six regions selected for study were characterized with respect to the following guidelines: (1) Potential repository media; (2) Quaternary tectonic conditions; (3) climatic change and geomorphic processes; (4) ground-water conditions; (5) ground-water quality; and (6) mineral and energy resources.The repository medium will function as the first natural barrier to radionuclide travel by virtue of associated slow ground-water velocity. The principal rock types considered as host media include granitic, intermediate, and mafic intrusive rocks; argillaceous rocks; salt and anhydrite; volcanic mudflow (laharic) breccias; some intrusive rhyolitic plugs and stocks; partially zeolitized tuff; and metamorphic rocks. In the unsaturated zone, the permeability and hydrologic properties of the rocks and the hydrologic setting are more important than the rock type. Media ideally should be permeable to provide drainage and should have a minimal water fluxThe ground-water flow path from a repository to the accessible environment needs to present major barriers to the transport of radionuclides. Factors considered in evaluating the ground-water conditions include ground-water traveltimes and quality, confining beds, and earth materials favorable for retardation of radionuclides. Ground-water velocities in the regions were calculated from estimated hydraulic properties of the rocks and gradients. Because site-specific data on hydraulic properties are not available, data from the literature were assembled and synthesized to obtain values for use in estimating ground-water velocities. Hydraulic conductivities for many rock types having granular and fracture permeability follow a log-normal distribution. Porosity for granular and very weathered

  14. Social impacts of hazardous and nuclear facilities and events: Implications for Nevada and the Yucca Mountain high-level nuclear waste repository; [Final report

    Energy Technology Data Exchange (ETDEWEB)

    Freudenburg, W.R. [Wisconsin Univ., Madison, WI (United States); Carter, L.F.; Willard, W. [Washington State Univ., Pullman, WA (United States); Lodwick, D.G. [Miami Univ., Oxford, OH (United States); Hardert, R.A. [Arizona State Univ., Tempe, AZ (United States); Levine, A.G. [State Univ. of New York, Buffalo, NY (United States). Dept. of Sociology; Kroll-Smith, S. [New Orleans Univ., LA (United States); Couch, S.R. [Pennsylvania State Univ., University Park, PA (United States); Edelstein, M.R. [Ramapo College, Mahwah, NJ (United States)

    1992-05-01

    Social impacts of a nuclear waste repository are described. Various case studies are cited such as Rocky Flats Plant, the Feed Materials Production Center, and Love Canal. The social impacts of toxic contamination, mitigating environmental stigma and loss of trust are also discussed.

  15. Workshop on the source term for radionuclide migration from high-level waste or spent nuclear fuel under realistic repository conditions: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, T.O.; Muller, A.B. (eds.)

    1985-07-01

    Sixteen papers were presented at the workshop. The fourteen full-length papers included in the proceedings were processed separately. Only abstracts were included for the following two papers: Data Requirements Based on Performance Assessment Analyses of Conceptual Waste Packages in Salt Repositories, and The Potential Effects of Radiation on the Source Term in a Salt Repository. (LM)

  16. Regulator's Workshop on The Role of Future Society and Biosphere in Demonstrating Compliance with High-Level Radioactive Waste Disposal Standards and Regulations

    Energy Technology Data Exchange (ETDEWEB)

    Avila, R. [Swedish Radiation Protection Authority, Stockholm (Sweden); Blommaert, W. [Agence Federale de Controle Nucleaire, Bruxelles (Belgium); Clark, R. [US Environmental Protection Agency (United States)] [and others

    2002-09-01

    This report summarizes the proceedings of a workshop, co-sponsored by the U.S. Environmental Protection Agency (US EPA) and the Swedish Radiation Protection Authority (SSI). The invitations to participate in the Workshop were primarily extended to authorities in countries with major nuclear waste programs involving geological disposal and using performance assessment methodology. The main objective of the Workshop was to develop a common understanding among regulators of the role of society and the biosphere in demonstrating compliance with regulations.

  17. Analysis of Surface Leaching Processes in Vitrified High-Level Nuclear Wastes Using In-Situ Raman Imaging and Atomistic Modeling - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, Joseph H.

    2001-04-24

    The in situ analysis of surface conditions of vitrified nuclear wastes can provide an important check of the burial status of radioactive objects without risk of radiation exposure. Raman spectroscopy was initially chosen as the most promising method for testing the surface conditions of glasses undergoing chemical corrosion, and was used extensively during the first year. However, it was determined that infrared reflection spectroscopy was better suited to this particular need and was used for the remaining two years to investigate the surface corrosion behavior of model silicate glasses for extension to nuclear waste glasses. The developed methodology is consistent with the known theory of optical propagation of dielectric media and uses the Kramers-Kronig formalism. The results show that it is possible to study the corrosion of glass by analyzing the glass surface using reflection fast Fourier infrared measurements and the newly developed ''dispersion analysis method.'' The data show how this analysis can be used to monitor the corrosion behavior of vitrified waste glasses over extended periods of storage.

  18. Japan-Australia Co-operative Program on research and development of technology for the management of high level radioactive wastes: phase II (1990-1995)

    Energy Technology Data Exchange (ETDEWEB)

    Banba, Tsunetaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hart, K.P. [eds.

    1996-05-01

    The major activities associated with Japan-Australia Co-operative Program were the preparation, characterization and subsequent testing of both Cm-doped Synroc containing PW-4b simulated waste and Cm-doped single-phase zirconolite and perovskite, and the initiation of studies on naturally-occurring zirconolites to study the long-term durability of this mineral phase over geological time. The preparation of the Cm-doped samples was carried out in JAERI`s WASTEF facility at Tokai, with technical information and assistance provided by ANSTO where necessary. The experiments were designed to induce accelerated radiation damage in Synroc samples that would correspond to periods of Synroc storage of up to 100,000 years. The results are of considerable importance in evaluating the potential of the Synroc process as a means of dealing with HLW waste streams and represent a significant contribution to the understanding of the ability of Synroc to immobilize HLW elements. Overall the Phase II Co-operative Program has continued the excellent co-operative working relationship between the staff at the two institutions, and provided a better understanding of the potential advantages and limitations of Synroc as a second generation waste form. The work has shown the need for additional studies to be carried out on the effect of the levels of Cm-doping on the Cm leach rate, extension of natural analogue studies to define the geological conditions under which zirconolite is stable and development of models to provide long-term predictions of releases of HLW elements from Synroc under a range of repository conditions. It is strongly recommended that the program carried out in Phase II of the Co-operative Agreement be extended for a further three years to allow additional information on the above areas to be collected and reported in a document providing an overview of the Co-operative Program and recommendations on HLW management strategies. (J.P.N.).

  19. The Properties of Calcined Anthracite

    OpenAIRE

    Holm, Katrine

    2017-01-01

    The aim of this thesis has been to look at the thermal behaviour of anthracite, coke and charcoal regarding their expansion during heating and their release of off-gases. Furthermore, to look at how the porosity and expansion of the anthracite is affected by the different temperatures during the calcining process. Two anthracites (A11 and A19), two cokes (C3 and C7) and one charcoal (BIO) were investigated. The porosity and surface area of A11 calcined at different temperatures were inves...

  20. Defense waste vitrification studies during FY 1980

    Energy Technology Data Exchange (ETDEWEB)

    Bjorklund, W.J.

    1981-08-01

    During FY-1980, Pacific Northwest Laboratory (PNL) tested three vitrification processes on simulated high-level radioactive waste typical of that stored or being produced at US defense facilities. Processes tested included a spray calciner/in-can melter, spray calciner/ceramic melter and direct liquid feeding of a ceramic melter. Tests were made on pilot-scale as well as fullscale equipment. Over 16,000 kg of glass product were produced from 68,000 L of simulated waste. Several compositions were tested, and the glass products were evaluated. Emphasis was placed on determining the processing rates and the ability of the waste to be processed. Off-gas data were collected on several runs. Major conclusions drawn from this test program are divided into processing results, glass-product results, and general information.

  1. On the importance of coupled THM processes to predict the long-term response of a generic salt repository for high-level nuclear waste

    Science.gov (United States)

    Blanco Martin, L.; Rutqvist, J.; Birkholzer, J. T.

    2013-12-01

    Salt is a potential medium for the underground disposal of nuclear waste because it has several assets, in particular its ability to creep and heal fractures generated by excavation and its water and gas tightness in the undisturbed state. In this research, we focus on disposal of heat-generating nuclear waste (such as spent fuel) and we consider a generic salt repository with in-drift emplacement of waste packages and subsequent backfill of the drifts with run-of-mine crushed salt. As the natural salt creeps, the crushed salt backfill gets progressively compacted and an engineered barrier system is subsequently created. In order to evaluate the integrity of the natural and engineered barriers over the long-term, it is important to consider the coupled effects of the thermal, hydraulic and mechanical processes that take place. In particular, the results obtained so far show how the porosity reduction of the crushed salt affects the saturation and pore pressure evolution throughout the repository, both in time and space. Such compaction is induced by the stress and temperature regime within the natural salt. Also, transport properties of the host rock are modified not only by thermo-mechanically and hydraulically-induced damaged processes, but also by healing/sealing of existing fractures. In addition, the THM properties of the backfill evolve towards those of the natural salt during the compaction process. All these changes are based on dedicated laboratory experiments and on theoretical considerations [1-3]. Different scenarios are modeled and compared to evaluate the relevance of different processes from the perspective of effective nuclear waste repositories. The sensitivity of the results to some parameters, such as capillarity, is also addressed. The simulations are conducted using an updated version of the TOUGH2-FLAC3D simulator, which is based on a sequential explicit method to couple flow and geomechanics [4]. A new capability for large strains and creep

  2. The siting record: An account of the programs of federal agencies and events that have led to the selection of a potential site for a geologic respository for high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Lomenick, T.F.

    1996-03-01

    This record of siting a geologic repository for high-level radioactive wastes (HLW) and spent fuel describes the many investigations that culminated on December 22, 1987 in the designation of Yucca Mountain (YM), as the site to undergo detailed geologic characterization. It recounts the important issues and events that have been instrumental in shaping the course of siting over the last three and one half decades. In this long task, which was initiated in 1954, more than 60 regions, areas, or sites involving nine different rock types have been investigated. This effort became sharply focused in 1983 with the identification of nine potentially suitable sites for the first repository. From these nine sites, five were subsequently nominated by the U.S. Department of Energy (DOE) as suitable for characterization and then, in 198