Sample records for nuclear waste technical

  1. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington are pleased to transmit a technical report prepared by the Nuclear Waste Technical Review Board (Board. Based on its review of data gathered by the DOE and the Center for Nuclear Waste Regulatory Analyses

  2. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    con144vf UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Dear Speaker Hastert, Senator Thurmond, and Secretary Richardson: The Nuclear Waste Technical Review with provisions of the Nuclear Waste Policy Amendments Act of 1987, Public Law 100-203. The Act requires the Board

  3. Appendix A Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    39 Appendices Appendices #12;Appendix A Nuclear Waste Technical Review Board Members: Curricula to the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon chairman on January 17, 1997, and Asia and on energy facility siting, including nuclear waste ship- ping and storage. In addition to his

  4. Appendix A Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    59 Appendices Appendices #12;Appendix A Nuclear Waste Technical Review Board Members: Curricula Cohon to serve on the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon, and Asia and on energy-facility siting, including nuclear waste shipping and storage. In addition to his

  5. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington.C. 20585 Dear Speaker Hastert, Senator Stevens, and Secretary Bodman: The Nuclear Waste Technical Review Board was created by Congress in the Nuclear Waste Policy Amendments Act (NWPAA) of 1987 and charged

  6. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    JEC187V3 UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 of Energy 1000 Independence Avenue, SW Washington, DC 20585 Dear Secretary O'Leary: At the Nuclear Waste UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington, VA

  7. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington Dear Speaker Pelosi, Senator Byrd, and Secretary Bodman: The Nuclear Waste Technical Review Board, and transporting high-level radioactive waste and spent nuclear fuel. The Board is required to report its findings

  8. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    of the overall system for managing spent fuel and defense high-level waste. The Board includes its findingsjlc029va UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300, D.C. 20585 Dear Speaker Gingrich, Senator Thurmond, and Secretary Peña: The Nuclear Waste Technical

  9. United States Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    Programs to Manage High-Level Radioactive Waste and Spent Nuclear Fuel in the United States and Other-Level Radioactive Waste and Spent Nuclear Fuel in the United States and Other Countries A Report to CongressUnited States Nuclear Waste Technical Review Board Experience Gained From Programs to Manage High

  10. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    's mission, continuing role, and refocused goals as the U.S. approach to managing spent nuclear fuel and high an ongoing and integrated technical peer review of all DOE activities related to managing spent nuclear fuel, and the characteristics and quantities of waste associated with adopting alternatives for spent nuclear fuel management

  11. U.S. Nuclear Waste Technical Review Board Performance Evaluation

    E-Print Network [OSTI]

    Addendum A Addendum A U.S. Nuclear Waste Technical Review Board Performance Evaluation Fiscal Year 2005 The U.S. Nuclear Waste Technical Review Board The Nuclear Waste Policy Amendments Act nuclear fuel and defense high-level radioactive waste. The Act also estab lished the U.S. Nuclear Waste

  12. U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to January to December 2000 The U.S. Congress are available at www.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE TECHNICAL REVIEW BOARD Dr. Jared L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 vii Table of Contents Appendices Appendix A Nuclear Waste Technical Review Board Members

  13. U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to The U.S. Congress And The Secretary of Energy January 1, 2003, to December 31, 2003 #12;U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to The U.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE TECHNICAL REVIEW BOARD 2003 Dr. Michael L. Corradini

  14. Appendix A U.S. Nuclear Waste Technical Review

    E-Print Network [OSTI]

    Appendices Appendices 37 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board as chair, on the U.S. Nuclear Regulatory Commission's Advisory Commit tee on Nuclear Waste. His areas to the Nuclear Waste Technical Review Board on June 26, 2002, by President George W. Bush. Dr. Abkowitz

  15. Nuclear Waste Technical Review Board Members Appendix A 53

    E-Print Network [OSTI]

    51 Appendix A Nuclear Waste Technical Review Board Members #12;#12;Appendix A 53 B. John Garrick, Ph.D., P.E. Chairman Dr. B. John Garrick was appointed to the U.S. Nuclear Waste Technical Review, on the U.S. Nuclear Regula- tory Commission's Advisory Committee on Nuclear Waste. His areas of expertise

  16. Appendix A U.S. Nuclear Waste Technical Review

    E-Print Network [OSTI]

    Appendices Appendices 31 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board.S. Nuclear Waste Technical Review Board as Chairman on September 10, 2004, by President George W. Bush. Dr­2004), 4 years as chair, on the U.S. Nuclear Regulatory Commission's Advisory Committee on Nuclear Waste

  17. U.S. Nuclear Waste Technical Review Board Members

    E-Print Network [OSTI]

    Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Jared L. Cohon, Ph.D.; Chairman On June 29, 1995, President Bill Clinton appointed Jared Cohon to the Nuclear Waste Technical, and Asia and on energy facil ity siting, including nuclear waste shipping and storage. In addition to his

  18. Appendix A U.S. Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    7 Appendices Appendices #12;Appendix A U.S. Nuclear Waste Technical Review Board Members: Curricula to the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon chairman on January 17, 1997, and Asia and on energy facility siting, including nuclear waste shipping and storage. In addition to his

  19. U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to January 1, 2001, to January 31, 2002 The U All NWTRB reports are available at www.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Appendices Appendix A U.S. Nuclear Waste Technical Review Board Members

  20. Nuclear Waste Technical Review Board Strategic Plan FY 20082013

    E-Print Network [OSTI]

    169 Appendix F Nuclear Waste Technical Review Board Strategic Plan FY 2008­2013 #12;#12;Appendix F 171 Nuclear Waste Technical Review Board Strategic Plan FY 2008­2013 suMMary sTaTeMenT of The Board T he Nuclear Waste Policy Amendments Act (NWPAA) of 1987 directed the U.S. Department of Energy (DOE

  1. U.S. Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    U.S. Nuclear Waste Technical Review Board Report to The U.S. Congressand The Secretary of Energy March 1, 2006­December 31, 2007 #12;#12;U.S. Nuclear Waste Technical Review Board Report to The U in this report. #12;#12;#12;U.S. Nuclear Waste Technical Review Board B. John Garrick, Ph.D., P.E., Chairman

  2. Other U.S. Nuclear Waste Technical Review Board Correspondence

    E-Print Network [OSTI]

    Appendix F Appendix F Other U.S. Nuclear Waste Technical Review Board Correspondence · Letter UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington, VA: Acknowledgement of letter · Letter from Robert R. Loux, Executive Director, Nevada Agency for Nuclear Projects

  3. Appendix A U.S. Nuclear Waste Technical Review

    E-Print Network [OSTI]

    Appendices Appendices 25 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Michael L. Corradini, Ph.D.; Chairman Dr. Michael L. Corradini was appointed to the Nuclear Waste and The Secretary of Energy Mark D. Abkowitz, Ph.D. Dr. Mark D. Abkowitz was appointed to the Nuclear Waste

  4. Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)

    E-Print Network [OSTI]

    .S. Department of Energy (DOE) activities related to the management of spent nuclear fuel (SNF) and high-cycle initiatives on the generation of spent nuclear fuel and high-level radioactive waste. The effort is led by Dr of the U.S. Nuclear Waste Technical Review Board's analysis of the impact of alternative fuel

  5. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    and issues related to the waste- management system, including transportation of spent nuclear fuel and high of Energy, including the possible development of a repository for disposing of spent nuclear fuel and high activities related to the possible disposal, packaging, and transportation of the country's spent nuclear

  6. U.S. Nuclear Waste Technical Review Board Correspondence with

    E-Print Network [OSTI]

    and the DOE's thermal-management strategy could result in excessive handling of spent-fuel assemblies as fuel system. The Board is concerned that assumptions related to receipt of spent fuel from utilitiesAppendix E Appendix E U.S. Nuclear Waste Technical Review Board Correspondence with U

  7. Communication Between the U.S. Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    Appendix F Appendix F Communication Between the U.S. Nuclear Waste Technical Review Board on Energy and Air Quality on March 25, 2004 143 #12;#12;Appendix F UNITED STATES NUCLEAR WASTE TECHNICAL much for your written questions related to my testimony on behalf of the Nuclear Waste Technical Review

  8. U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20042009

    E-Print Network [OSTI]

    Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2004­2009 (Revised March 2004) Statement of the Board The Nuclear Waste Policy Amendments Act of 1987 directed the U-level radioactive waste. The Act also established the U.S. Nuclear Waste Technical Review Board as an independent

  9. U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20042009

    E-Print Network [OSTI]

    Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2004­2009 (Revised March 2004) Statement of the Board The Nuclear Waste Policy Amendments Act of 1987 directed the U-level radioactive waste. The Act also established the U.S. Nuclear Waste Technical Review Board as an indepen dent

  10. U.S. Nuclear Waste Technical Review Board Fiscal Year 2002-2007 Strategic Plan

    E-Print Network [OSTI]

    Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Fiscal Year 2002-2007 Strategic Plan Statement of the Chairman The U.S. Nuclear Waste Technical Review Board was established of Energy Mission The Board's mission, established in the Nuclear Waste Policy Amendments Act (NWPAA

  11. Nuclear Waste Technical Review Board Strategic Plan for FY 2001-2006

    E-Print Network [OSTI]

    Appendix G Nuclear Waste Technical Review Board Strategic Plan for FY 2001-2006 (Revised March 2001) Statement of the Chairman The U.S. Nuclear Waste Technical Review Board was established as an independent agency of the United States Government on December 22, 1987, in the Nuclear Waste Policy Amendments Act

  12. U.S. Nuclear Waste Technical Review Board Jared L. Cohon, Ph.D.; Chairman

    E-Print Network [OSTI]

    Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Jared L. Cohon, Ph.D.; Chairman On June 29, 1995, President Bill Clinton appointed Jared Cohon to the Nuclear Waste Technical siting, including nuclear waste shipping and storage. In addition to his academic experience, he served

  13. Abbreviations and Acronyms Board U. S. Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    Tuff nonwelded unit SNF spent nuclear fuel SZ saturated zone SZEE saturated zone expert elicitation TCw of the Civilian Radioactive Waste Management Program." Presentation to Nuclear Waste Technical Review Board. MayAbbreviations and Acronyms Board U. S. Nuclear Waste Technical Review Board CFR Code of Federal

  14. U.S. NUclear WaSte techNical revieW Board

    E-Print Network [OSTI]

    technical context as important decisions are made on managing the nation's spent nuclear fuel and high, packaging, and transporting spent nuclear fuel and high-level radioactive waste is presented. The technicalU.S. NUclear WaSte techNical revieW Board Report to The U.S. Congress and The Secretary

  15. Nuclear Waste Technical Review Board Correspondence with the Department of Energy

    E-Print Network [OSTI]

    85 Appendix E Nuclear Waste Technical Review Board Correspondence with the Department of Energy #12;#12;Appendix E 87 Nuclear Waste Technical Review Board Correspondence with the Department of Energy I n of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The letters typically provide

  16. U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20032008

    E-Print Network [OSTI]

    Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2003­2008 Statement of the Chairman The Nuclear Waste Policy Amendments Act of 1987 directed the U.S. Department as the location of a permanent repository for dis posing of spent nuclear fuel and high-level radioactive waste

  17. The U.S. Nuclear Waste Technical Review Board evaluates the technical and scientific validity of ac-

    E-Print Network [OSTI]

    Chapter 1 Overview The U.S. Nuclear Waste Technical Review Board evaluates the technical in Nevada for its suit- ability as a location for a repository for high-level ra- dioactive waste (HLW) and spent nuclear fuel (SNF). The U.S. Department of Energy (DOE) began studying Yucca Mountain

  18. Nuclear Waste Technical Review Board Thermal-Response Evaluation of Yucca Mountain

    E-Print Network [OSTI]

    Nuclear Waste Technical Review Board Thermal-Response Evaluation of Yucca Mountain During of the thermal response of the proposed Yucca Mountain repository for various thermal loadings. The U. S. Nuclear Waste Technical Review Board (NWTRB) staff has developed calculation tools that allow performing

  19. Nuclear Waste Technical Review Board Members: Curricula Vitae

    E-Print Network [OSTI]

    mines, and storage projects, primarily in the fields of engineering geology and rock mechanics as an international consultant in the planning, designing, and construction of shafts, tunnels, dams, underground and Scisson and the U.S. Atomic Energy Commission on the design of underground openings for nuclear tests

  20. A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles

    E-Print Network [OSTI]

    Djokic, Denia

    2013-01-01T23:59:59.000Z

    m source, special nuclear, and waste   byproduct directly  United   States   Nuclear   Waste   Technical   Review  12,  2009.   NWPA  1982   Nuclear  Waste  Policy  Act  of  

  1. Corrosion-induced gas generation in a nuclear waste repository: Reactive geochemistry and multiphase flow effect

    E-Print Network [OSTI]

    Xu, T.

    2009-01-01T23:59:59.000Z

    Lying Repositories for Nuclear Waste, NAGRA Technical Reporthost rock formation for nuclear waste storage. EngineeringGas Generation in a Nuclear Waste Repository: Reactive

  2. Expected near-field thermal performance for nuclear waste repositories at potential salt sites: Technical report

    SciTech Connect (OSTI)

    McNulty, E.G.

    1987-08-01T23:59:59.000Z

    Thermal analyses were made for the environmental assessments of seven potential salt sites for a nuclear waste repository. These analyses predicted that potential repository sites in domal salts located in the Gulf Coast will experience higher temperature than those in bedded salts of Paradox and Palo Duro Basins, mainly because of higher ambient temperatures at depth. The TEMPV5 code, a semi-analytical heat transfer code for finite line sources, calculated temperatures for commercial high-level waste (CHLW) and spent fuel from pressurized-water reactors (SFPWR). Benchmarks with HEATING6, THAC-SIP-3D, STEALTH, and SPECTROM-41 showed that TEMPV5 agreed closely in the very near field around the waste package and approximately in the near-field and far-field regions of the repository. The analyses used site-specific thermal conductivities that were increased by 40% to compensate for reductions caused by testing technique, salt impurities, and other heterogeneities, and sampling disturbance. Analyses showed peak salt temperatures of 236/sup 0/C (CHLW) and 134/sup 0/C (SFPWR) for the bedded salt and 296/sup 0/C (CHLW) and 180/sup 0/C (SFPWR) for the domal salt. Analyses with uncorrected laboratory thermal conductivities would increase peak salt temperatures by about 120/sup 0/C for CHLW and about 60/sup 0/C for SFPWR. These temperature increases would increase the thermally induced flow of brine and accelerate corrosion of the waste package. 30 refs., 35 figs., 48 tabs.

  3. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    SciTech Connect (OSTI)

    Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

    1982-02-01T23:59:59.000Z

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

  4. Global Nuclear Energy Partnership Waste Treatment Baseline

    SciTech Connect (OSTI)

    Dirk Gombert; William Ebert; James Marra; Robert Jubin; John Vienna

    2008-05-01T23:59:59.000Z

    The Global Nuclear Energy Partnership program (GNEP) is designed to demonstrate a proliferation-resistant and sustainable integrated nuclear fuel cycle that can be commercialized and used internationally. Alternative stabilization concepts for byproducts and waste streams generated by fuel recycling processes were evaluated and a baseline of waste forms was recommended for the safe disposition of waste streams. Waste forms are recommended based on the demonstrated or expected commercial practicability and technical maturity of the processes needed to make the waste forms, and performance of the waste form materials when disposed. Significant issues remain in developing technologies to process some of the wastes into the recommended waste forms, and a detailed analysis of technology readiness and availability may lead to the choice of a different waste form than what is recommended herein. Evolving regulations could also affect the selection of waste forms.

  5. Nuclear waste solutions

    DOE Patents [OSTI]

    Walker, Darrel D. (1684 Partridge Dr., Aiken, SC 29801); Ebra, Martha A. (129 Hasty Rd., Aiken, SC 29801)

    1987-01-01T23:59:59.000Z

    High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  6. Net Zero Waste - Tools and Technical Support ...and other observations...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Net Zero Waste - Tools and Technical Support ...and other observations Net Zero Waste - Tools and Technical Support ...and other observations Presentation at Waste-to-Energy using...

  7. Technical requirements specification for tank waste retrieval

    SciTech Connect (OSTI)

    Lamberd, D.L.

    1996-09-26T23:59:59.000Z

    This document provides the technical requirements specification for the retrieval of waste from the underground storage tanks at the Hanford Site. All activities covered by this scope are conducted in support of the Tank Waste Remediation System (TWRS) mission.

  8. Waste transfer leaks technical basis document

    SciTech Connect (OSTI)

    ZIMMERMAN, B.D.

    2003-03-22T23:59:59.000Z

    This document provides technical support for the onsite radiological and toxicological, and offsite toxicological, portions of the waste transfer leak accident presented in the Documented Safety Analysis. It provides the technical basis for frequency and consequence bin selection, and selection of safety SSCs and TSRs.

  9. The Nuclear Waste Policy Act, as amended with appropriations acts appended

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    The Nuclear Waste Policy Act of 1982 provides for the development of repositories for the disposal of high-level radioactive waste and spent nuclear fuel, to establish a program of research, development and demonstration regarding the disposal of high-level radioactive waste and spent nuclear fuel. Titles 1 and 2 cover these subjects. Also included in this Act are: Title 3: Other provisions relating to radioactive waste; Title 4: Nuclear waste negotiation; Title 5: Nuclear waste technical review board; and Title 6: High-level radioactive waste. An appendix contains excerpts from appropriations acts from fiscal year 1984--1994.

  10. Development of Ceramic Waste Forms for High-Level Nuclear Waste Over the Last 30 Years

    SciTech Connect (OSTI)

    Vance, Eric [Institute of Materials and Engineering Science, Australian Nuclear Science and Technology Organisation, New Illawarra Road, Menai, NSW, 2234 (Australia)

    2007-07-01T23:59:59.000Z

    Many types of ceramics have been put forward for immobilisation of high-level waste (HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing some historical aspects of waste form research, the essential features of the chemical design and processing of these different ceramic types will be discussed briefly. Given acceptable laboratory and long-term predicted performance based on appropriately rigorous chemical design, the important processing parameters are mostly waste loading, waste throughput, footprint, offgas control/minimization, and the need for secondary waste treatment. It is concluded that the 'problem of high-level nuclear waste' is largely solved from a technical point of view, within the current regulatory framework, and that the main remaining question is which technical disposition method is optimum for a given waste. (author)

  11. Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility

    SciTech Connect (OSTI)

    Gates, R.; Glukhov, A.; Markowski, F.

    1996-06-01T23:59:59.000Z

    This technical report is a comprehensive evaluation of the proposal by the Ukrainian State Committee on Nuclear Power Utilization to create a central facility for radioactive waste (not spent fuel) processing. The central facility is intended to process liquid and solid radioactive wastes generated from all of the Ukrainian nuclear power plants and the waste generated as a result of Chernobyl 1, 2 and 3 decommissioning efforts. In addition, this report provides general information on the quantity and total activity of radioactive waste in the 30-km Zone and the Sarcophagus from the Chernobyl accident. Processing options are described that may ultimately be used in the long-term disposal of selected 30-km Zone and Sarcophagus wastes. A detailed report on the issues concerning the construction of a Ukrainian Central Radioactive Waste Processing Facility (CRWPF) from the Ukrainian Scientific Research and Design institute for Industrial Technology was obtained and incorporated into this report. This report outlines various processing options, their associated costs and construction schedules, which can be applied to solving the operating and decommissioning radioactive waste management problems in Ukraine. The costs and schedules are best estimates based upon the most current US industry practice and vendor information. This report focuses primarily on the handling and processing of what is defined in the US as low-level radioactive wastes.

  12. Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

    SciTech Connect (OSTI)

    Jacobson, Victor Levon

    2002-08-01T23:59:59.000Z

    U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodiumbearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements.

  13. affecting nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    influencing discussions on nuclear waste managementEthical Aspects of Nuclear Waste Alan Marshall Masaryk University Nuclear waste managementethical issues are just as important...

  14. Micro-Continuum Modeling of Nuclear Waste Glass Corrosion

    E-Print Network [OSTI]

    Steefel, Carl

    2014-01-01T23:59:59.000Z

    21. Grambow, B. (2006). Nuclear waste glasses – How durable?Continuum Modeling of Nuclear Waste Glass Corrosion AugustContinuum Modeling of Nuclear Waste Glass Corrosion Prepared

  15. The Social and Ethical Aspects of Nuclear Waste

    E-Print Network [OSTI]

    Marshall, Alan

    2005-01-01T23:59:59.000Z

    siting a high-level nuclear waste repository at Hanford,Eds. ), Public reactions to nuclear waste. Durham, NC: DukeInternational politics of nuclear waste. London: Macmillan.

  16. Waste management project technical baseline description

    SciTech Connect (OSTI)

    Sederburg, J.P.

    1997-08-13T23:59:59.000Z

    A systems engineering approach has been taken to describe the technical baseline under which the Waste Management Project is currently operating. The document contains a mission analysis, function analysis, requirement analysis, interface definitions, alternative analysis, system definition, documentation requirements, implementation definitions, and discussion of uncertainties facing the Project.

  17. Solid Waste Program technical baseline description

    SciTech Connect (OSTI)

    Carlson, A.B.

    1994-07-01T23:59:59.000Z

    The system engineering approach has been taken to describe the technical baseline under which the Solid Waste Program is currently operating. The document contains a mission analysis, function analysis, system definition, documentation requirements, facility and project bases, and uncertainties facing the program.

  18. WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008

    SciTech Connect (OSTI)

    Fellinger, A.

    2009-10-15T23:59:59.000Z

    The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

  19. Nuclear waste: our radioactive hot potato

    SciTech Connect (OSTI)

    Conselman, F.B.

    1984-01-01T23:59:59.000Z

    Nuclear industry inevitably produces nuclear waste, whose prudent, prompt and economic disposal is important to the national welfare. Technological problems of containment and isolation have apparently been solved. Underground or geologic disposal sites have the potential form permanent isolation, with salt, basalt, granite, shale, and tuff currently receiving principal attention as repository host rocks. Bedded salt deposits may offer the principal mechanical advantages, but in the northwestern United States the abundance of basalt at existing test sites has made it the subject of experimentation. However, psychological, political, and allegedly environmental obstructionism have stalled the process and virtually immobilized current construction. A program is suggested with the purpose of satisfying technical requirements for public protection while allaying the exaggerated fears of anti-nuclear factions.

  20. International nuclear waste management fact book

    SciTech Connect (OSTI)

    Abrahms, C W; Patridge, M D; Widrig, J E

    1995-11-01T23:59:59.000Z

    The International Nuclear Waste Management Fact Book has been compiled to provide current data on fuel cycle and waste management facilities, R and D programs, and key personnel in 24 countries, including the US; four multinational agencies; and 20 nuclear societies. This document, which is in its second year of publication supersedes the previously issued International Nuclear Fuel Cycle Fact Book (PNL-3594), which appeared annually for 12 years. The content has been updated to reflect current information. The Fact Book is organized as follows: National summaries--a section for each country that summarizes nuclear policy, describes organizational relationships, and provides addresses and names of key personnel and information on facilities. International agencies--a section for each of the international agencies that has significant fuel cycle involvement and a list of nuclear societies. Glossary--a list of abbreviations/acronyms of organizations, facilities, and technical and other terms. The national summaries, in addition to the data described above, feature a small map for each country and some general information that is presented from the perspective of the Fact Book user in the US.

  1. Technical Safety Requirements for the Waste Storage Facilities

    SciTech Connect (OSTI)

    Larson, H L

    2007-09-07T23:59:59.000Z

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 612 (A612) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2006). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., drum crushing, size reduction, and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A612 is located in the southeast quadrant of LLNL. The A612 fenceline is approximately 220 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A612 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting of buildings, tents, other structures, and open areas as described in Chapter 2 of the DSA. Section 2.4 of the DSA provides an overview of the buildings, structures, and areas in the WASTE STORAGE FACILITIES, including construction details such as basic floor plans, equipment layout, construction materials, controlling dimensions, and dimensions significant to the hazard and accident analysis. Chapter 5 of the DSA documents the derivation of the TSRs and develops the operational limits that protect the safety envelope defined for the WASTE STORAGE FACILITIES. This TSR document is applicable to the handling, storage, and treatment of hazardous waste, TRU WASTE, LLW, mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste received or generated in the WASTE STORAGE FACILITIES. Section 5, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the WASTE STORAGE FACILITIES. Programmatic Administrative Controls are in Section 5.6. This Introduction to the WASTE STORAGE FACILITIES TSRs is not part of the TSR limits or conditions and contains no requirements related to WASTE STORAGE FACILITIES operations or to the safety analyses of the DSA.

  2. Technical area status report for low-level mixed waste final waste forms. Volume 1

    SciTech Connect (OSTI)

    Mayberry, J.L.; DeWitt, L.M. [Science Applications International Corp., Idaho Falls, ID (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-08-01T23:59:59.000Z

    The Final Waste Forms (FWF) Technical Area Status Report (TASR) Working Group, the Vitrification Working Group (WG), and the Performance Standards Working Group were established as subgroups to the FWF Technical Support Group (TSG). The FWF TASR WG is comprised of technical representatives from most of the major DOE sites, the Nuclear Regulatory Commission (NRC), the EPA Office of Solid Waste, and the EPA`s Risk Reduction Engineering Laboratory (RREL). The primary activity of the FWF TASR Working Group was to investigate and report on the current status of FWFs for LLNM in this TASR. The FWF TASR Working Group determined the current status of the development of various waste forms described above by reviewing selected articles and technical reports, summarizing data, and establishing an initial set of FWF characteristics to be used in evaluating candidate FWFS; these characteristics are summarized in Section 2. After an initial review of available information, the FWF TASR Working Group chose to study the following groups of final waste forms: hydraulic cement, sulfur polymer cement, glass, ceramic, and organic binders. The organic binders included polyethylene, bitumen, vinyl ester styrene, epoxy, and urea formaldehyde. Section 3 provides a description of each final waste form. Based on the literature review, the gaps and deficiencies in information were summarized, and conclusions and recommendations were established. The information and data presented in this TASR are intended to assist the FWF Production and Assessment TSG in evaluating the Technical Task Plans (TTPs) submitted to DOE EM-50, and thus provide DOE with the necessary information for their FWF decision-making process. This FWF TASR will also assist the DOE and the MWIP in establishing the most acceptable final waste forms for the various LLMW streams stored at DOE facilities.

  3. Nuclear Materials: Reconsidering Wastes and Assets - 13193

    SciTech Connect (OSTI)

    Michalske, T.A. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable ('assets') to worthless ('wastes'). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or - in the case of high level waste - awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site's (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as 'waste' include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest. (authors)

  4. Technical Safety Requirements for the Waste Storage Facilities

    SciTech Connect (OSTI)

    Laycak, D T

    2008-06-16T23:59:59.000Z

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the 'Documented Safety Analysis for the Waste Storage Facilities' (DSA) (LLNL 2008). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting of buildings, tents, other structures, and open areas as described in Chapter 2 of the DSA. Section 2.4 of the DSA provides an overview of the buildings, structures, and areas in the WASTE STORAGE FACILITIES, including construction details such as basic floor plans, equipment layout, construction materials, controlling dimensions, and dimensions significant to the hazard and accident analysis. Chapter 5 of the DSA documents the derivation of the TSRs and develops the operational limits that protect the safety envelope defined for the WASTE STORAGE FACILITIES. This TSR document is applicable to the handling, storage, and treatment of hazardous waste, TRU WASTE, LLW, mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste received or generated in the WASTE STORAGE FACILITIES. Section 5, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the WASTE STORAGE FACILITIES. Programmatic Administrative Controls are in Section 5.6.

  5. Technical Safety Requirements for the Waste Storage Facilities

    SciTech Connect (OSTI)

    Laycak, D T

    2010-03-05T23:59:59.000Z

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2009). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting of buildings, tents, other structures, and open areas as described in Chapter 2 of the DSA. Section 2.4 of the DSA provides an overview of the buildings, structures, and areas in the WASTE STORAGE FACILITIES, including construction details such as basic floor plans, equipment layout, construction materials, controlling dimensions, and dimensions significant to the hazard and accident analysis. Chapter 5 of the DSA documents the derivation of the TSRs and develops the operational limits that protect the safety envelope defined for the WASTE STORAGE FACILITIES. This TSR document is applicable to the handling, storage, and treatment of hazardous waste, TRU WASTE, LLW, mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste received or generated in the WASTE STORAGE FACILITIES. Section 5, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the WASTE STORAGE FACILITIES. Programmatic Administrative Controls are in Section 5.4.

  6. BUOYANCY FLOW IN FRACTURES INTERSECTING A NUCLEAR WASTE REPOSITORY

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    discharge •side. As the wastes heat up the rock formationLBL—11112 "Heat Transfer to Nuclear Waste Disposal", ASMEv INTRODUCTION Heat released from a nuclear waste repository

  7. Hanford Waste Vitrification Plant technical manual

    SciTech Connect (OSTI)

    Larson, D.E. [ed.; Watrous, R.A.; Kruger, O.L. [and others

    1996-03-01T23:59:59.000Z

    A key element of the Hanford waste management strategy is the construction of a new facility, the Hanford Waste Vitrification Plant (HWVP), to vitrify existing and future liquid high-level waste produced by defense activities at the Hanford Site. The HWVP mission is to vitrify pretreated waste in borosilicate glass, cast the glass into stainless steel canisters, and store the canisters at the Hanford Site until they are shipped to a federal geological repository. The HWVP Technical Manual (Manual) documents the technical bases of the current HWVP process and provides a physical description of the related equipment and the plant. The immediate purpose of the document is to provide the technical bases for preparation of project baseline documents that will be used to direct the Title 1 and Title 2 design by the A/E, Fluor. The content of the Manual is organized in the following manner. Chapter 1.0 contains the background and context within which the HWVP was designed. Chapter 2.0 describes the site, plant, equipment and supporting services and provides the context for application of the process information in the Manual. Chapter 3.0 provides plant feed and product requirements, which are primary process bases for plant operation. Chapter 4.0 summarizes the technology for each plant process. Chapter 5.0 describes the engineering principles for designing major types of HWVP equipment. Chapter 6.0 describes the general safety aspects of the plant and process to assist in safe and prudent facility operation. Chapter 7.0 includes a description of the waste form qualification program and data. Chapter 8.0 indicates the current status of quality assurance requirements for the Manual. The Appendices provide data that are too extensive to be placed in the main text, such as extensive tables and sets of figures. The Manual is a revision of the 1987 version.

  8. Nuclear waste management. Semiannual progress report, October 1983-March 1984

    SciTech Connect (OSTI)

    McElroy, J.L.; Powell, J.A.

    1984-06-01T23:59:59.000Z

    Progress in the following studies on radioactive waste management is reported: defense waste technology; Nuclear Waste Materials Characterization Center; waste isolation; and supporting studies. 58 figures, 22 tables.

  9. RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    KOZLOWSKI, S.D.

    2007-05-30T23:59:59.000Z

    This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below.

  10. Evaluation of the Technical Basis for Extended Dry Storage and

    E-Print Network [OSTI]

    -- Executive Summary U.S. Nuclear Waste Technical Review Board December 2010 #12;U.S.U.S. Nuclear Waste Technical Review Board Authors This report was prepared for the U.S. Nuclear Waste Technical Review Board.NWTRB.GOV ii #12;Extended Dry Storage and Transportation of Used Nuclear Fuel U.S. Nuclear Waste Technical

  11. Plasma filtering techniques for nuclear waste remediation

    E-Print Network [OSTI]

    Gueroult, Renaud; Fisch, Nathaniel J

    2015-01-01T23:59:59.000Z

    Nuclear waste cleanup is challenged by the handling of feed stocks that are both unknown and complex. Plasma filtering, operating on dissociated elements, offers advantages over chemical methods in processing such wastes. The costs incurred by plasma mass filtering for nuclear waste pretreatment, before ultimate disposal, are similar to those for chemical pretreatment. However, significant savings might be achieved in minimizing the waste mass. This advantage may be realized over a large range of chemical waste compositions, thereby addressing the heterogeneity of legacy nuclear waste.

  12. Bubblers Speed Nuclear Waste Processing at SRS

    SciTech Connect (OSTI)

    None

    2010-11-14T23:59:59.000Z

    At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

  13. Bubblers Speed Nuclear Waste Processing at SRS

    ScienceCinema (OSTI)

    None

    2014-08-06T23:59:59.000Z

    At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

  14. Technical basis for classification of low-activity waste fraction from Hanford site tanks

    SciTech Connect (OSTI)

    Petersen, C.A.

    1996-09-20T23:59:59.000Z

    The overall objective of this report is to provide a technical basis to support a U.S. Nuclear Regulatory Commission determination to classify the low-activity waste from the Hanford Site single-shell and double-shell tanks as `incidental` wastes after removal of additional radionuclides and immobilization.The proposed processing method, in addition to the previous radionuclide removal efforts, will remove the largest practical amount of total site radioactivity, attributable to high-level waste, for disposal is a deep geologic repository. The remainder of the waste would be considered `incidental` waste and could be disposed onsite.

  15. Geotechnical support and topical studies for nuclear waste geologic repositories

    SciTech Connect (OSTI)

    Not Available

    1989-01-01T23:59:59.000Z

    The present report lists the technical reviews and comments made during the fiscal year 1988 and summarizes the technical progress of the topical studies. In the area of technical assistance, there were numerous activities detailed in the next section. These included 24 geotechnical support activities, including reviews of 6 Study Plans (SP) and participation in 6 SP Review Workshops, review of one whole document Site Characterization Plan (SCP) and participation in the Assembled Document SCP Review Workshops by 6 LBL reviewers; the hosting of a DOE program review, the rewriting of the project statement of work, 2 trips to technical and planning meetings; preparation of proposed work statements for two new topics for DOE, and 5 instances of technical assistance to DOE. These activities are described in a Table in the following section entitled Geoscience Technical Support for Nuclear Waste Geologic Repositories.''

  16. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

    Duffy, C.J.; Ogard, A.E.

    1982-02-01T23:59:59.000Z

    Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

  17. BUOYANCY FLOW IN FRACTURES INTERSECTING A NUCLEAR WASTE REPOSITORY

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    11112 "Heat Transfer to Nuclear Waste Disposal", ASME WinterIN FPACTUHES INTERSECTING A NUCLEAR WASTE REPOSITORY J.S.Y.Heat released from a nuclear waste repository in a

  18. INVESTIGATIONS IN GRANITE AT STRIPA, SWEDEN FOR NUCLEAR WASTE STORAGE

    E-Print Network [OSTI]

    Witherspoon, P.A.

    2010-01-01T23:59:59.000Z

    STRIPA, SWEDEN FOR NUCLEAR WASTE STORAGE P. A. tfitherspoon,GRANITE AT STRIPA, SWEDEN FOR NUCLEAR WASTE STORAGE by P. A.Final and safe storage of nuclear waste materials is one of

  19. Nuclear Waste and the Distant Future Nuclear Waste and the Distant Future

    E-Print Network [OSTI]

    Kammen, Daniel M.

    Nuclear Waste and the Distant Future 1 Nuclear Waste and the Distant Future PER F. PETERSON WILLIAM://www.issues.org/22.4/peterson.html Regulation of nuclear hazards must be consistent with rules governing other of the radioactive material generated by nuclear energy decays away over short times ranging from minutes to several

  20. The Initial Environmental Assessments for the Nuclear Waste Repository under Section 112 of the Nuclear Waste Policy Act

    E-Print Network [OSTI]

    Montange, Charles H.

    1985-01-01T23:59:59.000Z

    NWPA. See note 139 supra. NUCLEAR WASTE ASSESSMENTS cess orof the program. If the Nuclear Waste Policy Act is to work,potentially acceptable for nuclear waste repositories was

  1. Nuclear Waste Management. Semiannual progress report, October 1984-March 1985

    SciTech Connect (OSTI)

    McElroy, J.L.; Powell, J.A. (comps.)

    1985-06-01T23:59:59.000Z

    Progress reports are presented for the following studies on radioactive waste management: defense waste technology; nuclear waste materials characterization center; and supporting studies. 19 figs., 29 tabs.

  2. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    thermohydroiogic behavior of nuclear waste r e p o s i t o rground repository for nuclear wastes in hard r o d ' .RELATED PROBLEMS IN A NUCLEAR WASTE REPOSITORY T h i s b i b

  3. Salt disposal of heat-generating nuclear waste.

    SciTech Connect (OSTI)

    Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

    2011-01-01T23:59:59.000Z

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United States repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, helps define a clear strategy for a heat-generating nuclear waste repository in salt.

  4. Introduction to Nuclear Waste Management Nuclear Waste is a type of radioactive waste that is usually the by-product of

    E-Print Network [OSTI]

    Auerbach, Scott M.

    Introduction to Nuclear Waste Management Nuclear Waste is a type of radioactive waste Meltdowns Bad? - Nuclear Fallout -Water Pollution - Human Health Nuclear Waste Management The following examples are from our own exploration of the impact of nuclear waste... Brainstorm: What Do You

  5. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-26T23:59:59.000Z

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  6. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-25T23:59:59.000Z

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  7. E-Print Network 3.0 - actinide waste forms Sample Search Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    separated into a high-level radioactive stream and a fission product gases waste stream... Nuclear Waste Assessment System for Technical ... Source: U.S. Nuclear Waste...

  8. Impact assessment of draft DOE Order 5820.2B. Radioactive Waste Technical Support Program

    SciTech Connect (OSTI)

    NONE

    1995-04-01T23:59:59.000Z

    The Department of Energy (DOE) has prepared a revision to DOE Order 5820.2A, entitled ``Radioactive Waste Management.`` DOE issued DOE Order 5820.2A in September 1988 and, as the title implies, it covered only radioactive waste forms. The proposed draft order, entitled ``Waste Management,`` addresses the management of both radioactive and nonradioactive waste forms. It also includes spent nuclear fuel, which DOE does not consider a waste. Waste forms covered include hazardous waste, high-level waste, transuranic (TRU) waste, low-level radioactive waste, uranium and thorium mill tailings, mixed waste, and sanitary waste. The Radioactive Waste Technical Support Program (TSP) of Leached Idaho Technologies Company (LITCO) is facilitating the revision of this order. The EM Regulatory Compliance Division (EM-331) has requested that TSP estimate the impacts and costs of compliance with the revised order. TSP requested Dames & Moore to aid in this assessment by comparing requirements in Draft Order 5820.2B to ones in DOE Order 5820.2A and other DOE orders and Federal regulations. The assessment started with a draft version of 5820.2B dated January 14, 1994. DOE has released three updated versions of the draft order since then (dated May 20, 1994; August 26, 1994; and January 23, 1995). Each time DOE revised the order, Dames and Moore updated the assessment work to reflect the text changes. This report reflects the January 23, 1995 version of the draft order.

  9. Removing nuclear waste, one shipment at a time

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Stories Removing nuclear waste, one shipment at a time Removing nuclear waste, one shipment at a time The Lab's 1,000th shipment of transuranic waste recently left Los Alamos,...

  10. Review of Used Nuclear Fuel Storage and Transportation Technical...

    Broader source: Energy.gov (indexed) [DOE]

    action based on the comparison. Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis More Documents & Publications Review of Used Nuclear Fuel...

  11. Natural analogues of nuclear waste glass corrosion.

    SciTech Connect (OSTI)

    Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

    1999-01-06T23:59:59.000Z

    This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses.

  12. Doing the impossible: Recycling nuclear waste

    ScienceCinema (OSTI)

    None

    2013-04-19T23:59:59.000Z

    A Science Channel feature explores how Argonne techniques could be used to safely reduce the amount of radioactive waste generated by nuclear power?the most plentiful carbon-neutral energy source. Read more at http://www.anl.gov/Media_Center/ArgonneNow/Fall_2009/nuclear.html

  13. Waste Minimization Policy at the Romanian Nuclear Power Plant

    SciTech Connect (OSTI)

    Andrei, V.; Daian, I.

    2002-02-26T23:59:59.000Z

    The radioactive waste management system at Cernavoda Nuclear Power Plant (NPP) in Romania was designed to maintain acceptable levels of safety for workers and to protect human health and the environment from exposure to unacceptable levels of radiation. In accordance with terminology of the International Atomic Energy Agency (IAEA), this system consists of the ''pretreatment'' of solid and organic liquid radioactive waste, which may include part or all of the following activities: collection, handling, volume reduction (by an in-drum compactor, if appropriate), and storage. Gaseous and aqueous liquid wastes are managed according to the ''dilute and discharge'' strategy. Taking into account the fact that treatment/conditioning and disposal technologies are still not established, waste minimization at the source is a priority environmental management objective, while waste minimization at the disposal stage is presently just a theoretical requirement for future adopted technologies . The necessary operational and maintenance procedures are in place at Cernavoda to minimize the production and contamination of waste. Administrative and technical measures are established to minimize waste volumes. Thus, an annual environmental target of a maximum 30 m3 of radioactive waste volume arising from operation and maintenance has been established. Within the first five years of operations at Cernavoda NPP, this target has been met. The successful implementation of the waste minimization policy has been accompanied by a cost reduction while the occupational doses for plant workers have been maintained at as low as reasonably practicable levels. This paper will describe key features of the waste management system along with the actual experience that has been realized with respect to minimizing the waste volumes at the Cernavoda NPP.

  14. Technical area status report for waste destruction and stabilization

    SciTech Connect (OSTI)

    Dalton, J.D.; Harris, T.L.; DeWitt, L.M. [Science Applications International Corp., Idaho Falls, ID (United States)

    1993-08-01T23:59:59.000Z

    The Office of Environmental Restoration and Waste Management (EM) was established by the Department of Energy (DOE) to direct and coordinate waste management and site remediation programs/activities throughout the DOE complex. In order to successfully achieve the goal of properly managing waste and the cleanup of the DOE sites, the EM was divided into five organizations: the Office of Planning and Resource Management (EM-10); the Office of Environmental Quality Assurance and Resource Management (EM-20); the Office of Waste Operations (EM-30); the Office of Environmental Restoration (EM-40); and the Office of Technology and Development (EM-50). The mission of the Office of Technology Development (OTD) is to develop treatment technologies for DOE`s operational and environmental restoration wastes where current treatment technologies are inadequate or not available. The Mixed Waste Integrated Program (MWIP) was created by OTD to assist in the development of treatment technologies for the DOE mixed low-level wastes (MLLW). The MWIP has established five Technical Support Groups (TSGs) whose purpose is to identify, evaluate, and develop treatment technologies within five general technical areas representing waste treatment functions from initial waste handling through generation of final waste forms. These TSGs are: (1) Front-End Waste Handling, (2) Physical/Chemical Treatment, (3) Waste Destruction and Stabilization, (4) Second-Stage Destruction and Offgas Treatment, and (5) Final Waste Forms. This report describes the functions of the Waste Destruction and Stabilization (WDS) group. Specifically, the following items are discussed: DOE waste stream identification; summary of previous efforts; summary of WDS treatment technologies; currently funded WDS activities; and recommendations for future activities.

  15. Technical strategy for the management of INEEL spent nuclear fuel

    SciTech Connect (OSTI)

    NONE

    1997-03-01T23:59:59.000Z

    This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

  16. THERMODYNAMIC TABLES FOR NUCLEAR WASTE ISOLATION, V.1: AQUEOUS SOLUTIONS DATABASE

    E-Print Network [OSTI]

    Phillips, S.L.; Hale, F.V.; Silvester, L.F.

    2008-01-01T23:59:59.000Z

    National Laboratory Nuclear Waste Nanagement Division Upton,~ermodynamic Tables for Nuclear Waste Isolation Vol. I. Aq~Thermodynamic Tables for Nuclear Waste Isolation. Vol 1.

  17. Water-Steel Canister Interaction and H2 Gas Pressure Buildup in a Nuclear Waste Repository

    E-Print Network [OSTI]

    Xu, Tianfu; Senger, Rainer; Finstele, Stefan

    2008-01-01T23:59:59.000Z

    Nuclear Waste Repository T. Xu & S. Finsteiie Earth Sciencesdeep lying repositories for nuclear waste. Nagra Techni­ calthe system state in a nuclear waste re­ pository. 2 PROCESS

  18. GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    Interactions between nuclear waste and surrounding rock.AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION INwill provide Office of Nuclear Waste Isolation and the

  19. Nuclear waste incineration technology status

    SciTech Connect (OSTI)

    Ziegler, D.L.; Lehmkuhl, G.D.; Meile, L.J.

    1981-07-15T23:59:59.000Z

    The incinerators developed and/or used for radioactive waste combustion are discussed and suggestions are made for uses of incineration in radioactive waste management programs and for incinerators best suited for specific applications. Information on the amounts and types of radioactive wastes are included to indicate the scope of combustible wastes being generated and in existence. An analysis of recently developed radwaste incinerators is given to help those interested in choosing incinerators for specific applications. Operating information on US and foreign incinerators is also included to provide additional background information. Development needs are identified for extending incinerator applications and for establishing commercial acceptance.

  20. Energy Department and Catholic University Improve Safety of Nuclear Waste

    Broader source: Energy.gov [DOE]

    A new waste processing plant in Washington will help to safely remove nuclear and chemical waste, thanks to research from Catholic University.

  1. Annual report on the development and characterization of solidified forms for nuclear wastes, 1979

    SciTech Connect (OSTI)

    Chick, L.A.; McVay, G.L.; Mellinger, G.B.; Roberts, F.P.

    1980-12-01T23:59:59.000Z

    Development and characterization of solidified nuclear waste forms is a major continuing effort at Pacific Northwest Laboratory. Contributions from seven programs directed at understanding chemical composition, process conditions, and long-term behaviors of various nuclear waste forms are included in this report. The major findings of the report are included in extended figure captions that can be read as brief technical summaries of the research, with additional information included in a traditional narrative format. Waste form development proceeded on crystalline and glass materials for high-level and transuranic (TRU) wastes. Leaching studies emphasized new areas of research aimed at more basic understanding of waste form/aqueous solution interactions. Phase behavior and thermal effects research included studies on crystal phases in defense and TRU waste glasses and on liquid-liquid phase separation in borosilicate waste glasses. Radiation damage effects in crystals and glasses from alpha decay and from transmutation are reported.

  2. The Potential of Cellulosic Ethanol Production from Municipal Solid Waste: A Technical and Economic Evaluation

    E-Print Network [OSTI]

    Shi, Jian; Ebrik, Mirvat; Yang, Bin; Wyman, Charles E.

    2009-01-01T23:59:59.000Z

    Waste: A Technical and Economic Evaluation" Jian Shi, Mirvatwaste: A Technical and Economic Evaluation Jian Shi, MirvatIn addition, techno- economic evaluation of large scale

  3. Nuclear waste management. Quarterly progress report, January-March 1980

    SciTech Connect (OSTI)

    Platt, A.M.; Powell, J.A. (comps.)

    1980-06-01T23:59:59.000Z

    Reported are: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, criteria for defining waste isolation, and spent fuel and pool component integrity. (DLC)

  4. Technical Area (TA)-54 Area G Nitrate-Salt Waste Container Response...

    Office of Environmental Management (EM)

    Technical Area (TA)-54 Area G Nitrate-Salt Waste Container Response Instructions Technical Area (TA)-54 Area G Nitrate-Salt Waste Container Response Instructions This document was...

  5. Recovery of fissile materials from nuclear wastes

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1999-01-01T23:59:59.000Z

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

  6. Scientific Solutions to Nuclear Waste Environmental Challenges

    SciTech Connect (OSTI)

    Johnson, Bradley R.

    2014-01-30T23:59:59.000Z

    The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount of time it takes for one-half of the material to undergo radioactive decay.) In general, the ideal material would need to be durable for approximately 10 half-lives to allow the activity to decay to negligible levels. However, the potential health effects of each radionuclide vary depending on what type of radiation is emitted, the energy of that emission, and the susceptibility for the human body to accumulate and concentrate that particular element. Consequently, actual standards tend to be based on limiting the dose (energy deposited per unit mass) that is introduced into the environment. The Environmental Protection Agency (EPA) has the responsibility to establish standards for nuclear waste disposal to protect the health and safety of the public. For example, the Energy Policy Act of 1992 directed the EPA to establish radiation protection standards for the Yucca Mountain geologic repository for nuclear wastes. The standards for Yucca Mountain were promulgated in 2008, and limit the dose to 15 millirem per year for the first 10,000 years, and 100 milirem per year between 10,000 years and 1 million years (40 CFR Part 197; http://www.epa.gov/radiation/yucca/2008factsheet.html). So, the challenge is two-fold: (1) develop a material (a waste form) that is capable of immobilizing the waste over geologic time scales, and (2) develop a process to convert the radioactive sludge in the tanks into this durable waste form material. Glass: Hard, durable, inert, and with infinite chemical versatility Molten glass is a powerful solvent liquid, which can be designed to dissolve almost anything. When solidified, it can be one of the most chemically inert substances known to man. Nature's most famous analogue to glass is obsidian, a vitreous product of volcanic activity; formations over 17 million years old have been found. Archaeologists have found man-made glass specimens that are five thousand years old.

  7. Nuclear waste management. Semiannual progress report, October 1982-March 1983

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1983-06-01T23:59:59.000Z

    This document is one of a series of technical progress reports designed to report radioactive waste management programs at the Pacific Northwest Laboratory. Accomplishments in the following programs are reported: waste stabilization; Materials Characterization Center; waste isolation; low-level waste management; remedial action; and supporting studies.

  8. Nuclear Waste Fund fee adequacy: An assessment

    SciTech Connect (OSTI)

    NONE

    1990-11-01T23:59:59.000Z

    The purpose of this report is to present the Department of Energy`s (the Department) analysis of the adequacy of the 1.00 mill per kilowatt-hour (kWh) fee being paid by the utilities generating nuclear power for the permanent disposal of their spent nuclear fuel (SNF). In accordance with the Nuclear Waste Policy Act (NWPA), the SNF would be disposed of in a geologic repository to be developed by the Department. An annual analysis of the fee`s adequacy is required by the NWPA.

  9. USED NUCLEAR MATERIALS AT SAVANNAH RIVER SITE: ASSET OR WASTE?

    SciTech Connect (OSTI)

    Magoulas, V.

    2013-06-03T23:59:59.000Z

    The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable (“assets”) to worthless (“wastes”). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or – in the case of high level waste – awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site’s (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as “waste” include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest.

  10. National briefing summaries: Nuclear fuel cycle and waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1988-12-01T23:59:59.000Z

    The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

  11. A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles

    E-Print Network [OSTI]

    Djokic, Denia

    2013-01-01T23:59:59.000Z

    Spent  Nuclear   Fuel,”   Integrated   Radioactive   Waste   Management  spent  nuclear  fuel”  [42  USC  10101]   as   high-­?level   waste   potentially   neglects   the   waste   management  

  12. HEAT TRANSFER ANALYSIS FOR NUCLEAR WASTE SOLIDIFICATION CONTAINER

    SciTech Connect (OSTI)

    Lee, S.

    2009-06-01T23:59:59.000Z

    The Nuclear Nonproliferation Programs Design Authority is in the design stage of the Waste Solidification Building (WSB) for the treatment and solidification of the radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility (PDCF) and Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The waste streams will be mixed with a cementitious dry mix in a 55-gallon waste container. Savannah River National Laboratory (SRNL) has been performing the testing and evaluations to support technical decisions for the WSB. Engineering Modeling & Simulation Group was requested to evaluate the thermal performance of the 55-gallon drum containing hydration heat source associated with the current baseline cement waste form. A transient axi-symmetric heat transfer model for the drum partially filled with waste form cement has been developed and heat transfer calculations performed for the baseline design configurations. For this case, 65 percent of the drum volume was assumed to be filled with the waste form, which has transient hydration heat source, as one of the baseline conditions. A series of modeling calculations has been performed using a computational heat transfer approach. The baseline modeling results show that the time to reach the maximum temperature of the 65 percent filled drum is about 32 hours when a 43 C initial cement temperature is assumed to be cooled by natural convection with 27 C external air. In addition, the results computed by the present model were compared with analytical solutions. The modeling results will be benchmarked against the prototypic test results. The verified model will be used for the evaluation of the thermal performance for the WSB drum. Detailed results and the cases considered in the calculations will be discussed here.

  13. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    NONE

    1995-09-01T23:59:59.000Z

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  14. Technical specifications for mechanical recycling of agricultural plastic waste

    SciTech Connect (OSTI)

    Briassoulis, D., E-mail: briassou@aua.gr; Hiskakis, M.; Babou, E.

    2013-06-15T23:59:59.000Z

    Highlights: • Technical specifications for agricultural plastic wastes (APWs) recycling proposed. • Specifications are the base for best economical and environmental APW valorisation. • Analysis of APW reveals inherent characteristics and constraints of APW streams. • Thorough survey on mechanical recycling processes and industry as it applies to APW. • Specifications for APW recycling tested, adjusted and verified through pilot trials. - Abstract: Technical specifications appropriate for the recycling of agricultural plastic wastes (APWs), widely accepted by the recycling industry were developed. The specifications establish quality standards to be met by the agricultural plastics producers, users and the agricultural plastic waste management chain. They constitute the base for the best economical and environmental valorisation of the APW. The analysis of the APW streams conducted across Europe in the framework of the European project “LabelAgriWaste” revealed the inherent characteristics of the APW streams and the inherent constraints (technical or economical) of the APW. The APW stream properties related to its recycling potential and measured during pilot trials are presented and a subsequent universally accepted simplified and expanded list of APW recycling technical specifications is proposed and justified. The list includes two sets of specifications, applied to two different quality categories of recyclable APW: one for pellet production process (“Quality I”) and another one for plastic profile production process (“Quality II”). Parameters that are taken into consideration in the specifications include the APW physical characteristics, contamination, composition and degradation. The proposed specifications are focused on polyethylene based APW that represents the vast majority of the APW stream. However, the specifications can be adjusted to cover also APW of different materials (e.g. PP or PVC) that are found in very small quantities in protected cultivations in Europe. The adoption of the proposed specifications could transform this waste stream into a labelled commodity traded freely in the market and will constitute the base for the best economical and environmental valorisation of the APW.

  15. TECHNICAL ADVANCEMENTS AND ISSUES ASSOCIATED WITH

    E-Print Network [OSTI]

    of Energy United States Nuclear Waste Technical Review Board #12;#12;U.S. Nuclear Waste Technical Review in the report are those of the Board and not necessarily those of the reviewers. #12;U.S. NUCLEAr WASTE TECh.D., P.E. Duke University Durham, North Carolina #12;NUCLEAr WASTE TEChNiCAL rEviEW BOArD STAff EXECUTIVE

  16. Advanced pyrochemical technologies for minimizing nuclear waste

    SciTech Connect (OSTI)

    Bronson, M.C.; Dodson, K.E.; Riley, D.C.

    1994-06-01T23:59:59.000Z

    The Department of Energy (DOE) is seeking to reduce the size of the current nuclear weapons complex and consequently minimize operating costs. To meet this DOE objective, the national laboratories have been asked to develop advanced technologies that take uranium and plutonium, from retired weapons and prepare it for new weapons, long-term storage, and/or final disposition. Current pyrochemical processes generate residue salts and ceramic wastes that require aqueous processing to remove and recover the actinides. However, the aqueous treatment of these residues generates an estimated 100 liters of acidic transuranic (TRU) waste per kilogram of plutonium in the residue. Lawrence Livermore National Laboratory (LLNL) is developing pyrochemical techniques to eliminate, minimize, or more efficiently treat these residue streams. This paper will present technologies being developed at LLNL on advanced materials for actinide containment, reactors that minimize residues, and pyrochemical processes that remove actinides from waste salts.

  17. Seal welded cast iron nuclear waste container

    DOE Patents [OSTI]

    Filippi, Arthur M. (Pittsburgh, PA); Sprecace, Richard P. (Murrysville, PA)

    1987-01-01T23:59:59.000Z

    This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

  18. Nuclear waste programs; Semiannual progress report, October 1991--March 1992

    SciTech Connect (OSTI)

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Finn, P.A.; Gerding, T.J.; Hoh, J.C. [and others

    1993-11-01T23:59:59.000Z

    This document reports on the work done by the Nuclear Waste Programs of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1991-March 1992. In these programs, studies are underway on the performance of waste glass and spent fuel in projected nuclear repository conditions to provide input to the licensing of the nation`s high-level waste repositories

  19. The necessity for permanence : making a nuclear waste storage facility

    E-Print Network [OSTI]

    Stupay, Robert Irving

    1991-01-01T23:59:59.000Z

    The United States Department of Energy is proposing to build a nuclear waste storage facility in southern Nevada. This facility will be designed to last 10,000 years. It must prevent the waste from contaminating the ...

  20. MIXING OF INCOMPATIBLE MATERIALS IN WASTE TANKS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    SANDGREN, K.R.

    2003-10-15T23:59:59.000Z

    This document presents onsite radiological, onsite toxicological, and offsite toxicological consequences, risk binning, and control decision results for the mixing of incompatible materials in waste tanks representative accident. This technical basis document was developed to support the tank farms documented safety analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  1. Extraction of cesium and strontium from nuclear waste

    DOE Patents [OSTI]

    Davis, Jr., Milton W. (Lexington, SC); Bowers, Jr., Charles B. (Columbia, SC)

    1988-01-01T23:59:59.000Z

    Cesium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5) [1-hydroxy-2-ethylhexyl]benzo 18-crown-6 compound and a cation exchanger in a matrix solution. Strontium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5') [1-hydroxyheptyl]cyclohexo 18-crown-6 compound, and a cation exchanger in a matrix solution.

  2. Extraction of cesium and strontium from nuclear waste

    DOE Patents [OSTI]

    Davis, M.W. Jr.; Bowers, C.B. Jr.

    1988-06-07T23:59:59.000Z

    Cesium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4[prime](5) [1-hydroxy-2-ethylhexyl]benzo 18-crown-6 compound and a cation exchanger in a matrix solution. Strontium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4[prime](5[prime]) [1-hydroxyheptyl]cyclohexo 18-crown-6 compound, and a cation exchanger in a matrix solution. 3 figs.

  3. The Italian Activities in the Field of Nuclear Waste Management - 12439

    SciTech Connect (OSTI)

    Giorgiantoni, Giorgio; Marzo, Giuseppe A.; Sepielli, Massimo [ENEA, C. R. Casaccia, Roma (Italy)

    2012-07-01T23:59:59.000Z

    The Italian situation in the field of nuclear waste management is characterized by a relative small quantity of wastes, as a consequence of the giving up of energy production by nuclear generation in 1986. Notwithstanding this situation, Italy is a unique case study since the country needs to undertake the final decommissioning of four shut-down NPPs (size 100-200 MWe), each one different from the others. Therefore all the regulatory, technical, and financial actions are needed in the same way as if there was actual nuclear generation. Furthermore, the various non-power generating applications of nuclear energy still require management, a legal framework, a regulatory body, an industrial structure, and technical know-how. Notwithstanding the absence of energy production from nuclear sources, the country has the burden of radioactive waste management from the previous nuclear operations, which obliges it to implement at first a robust legislative framework, then to explore all the complex procedures to achieve the localization of the national interim storage facility, not excluding the chance to have a European regional facility for geologic disposal, under the clauses of the Council Directive of 19 July 2011 'Establishing a Community Framework for the Responsible and Safe Management of Radioactive Waste'. Then, as far as industrial, medical and R and D aspects, the improvement of the legislative picture, the creation of a regulatory body, is a good start for the future, to achieve the best efficiency of the Italian system. (authors)

  4. Waste Stream Analyses for Nuclear Fuel Cycles

    SciTech Connect (OSTI)

    N. R. Soelberg

    2010-08-01T23:59:59.000Z

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  5. A comparative simulation study of coupled THM processes and their effect on fractured rock permeability around nuclear waste repositories

    E-Print Network [OSTI]

    Rutqvist, Jonny

    2008-01-01T23:59:59.000Z

    hydrothermomechanical design of nuclear waste repositories.Associated with Nuclear Waste Repositories, Academic Press,rock permeability around nuclear waste repositories Jonny

  6. Multiple-code simulation study of the long-term EDZ evolution of geological nuclear waste repositories

    E-Print Network [OSTI]

    Rutqvist, J.

    2008-01-01T23:59:59.000Z

    Associated with Nuclear Waste Repositories, Academic Press,C (eds) Rock Mechanics of Nuclear Waste Repositories, Veil,EDZ Evolution of Geological Nuclear Waste Repositories Jonny

  7. Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2007-08-28T23:59:59.000Z

    The order establishes Central Technical Authority and Chief of Nuclear Safety/Chief of Defense Nuclear Safety responsibilities and requirements directed by the Secretary of Energy in the development and issuance of Department of Energy regulations and directives that affect nuclear safety. No cancellations.

  8. Greater-than-Class C low-level radioactive waste characterization. Appendix E-5: Impact of the 1993 NRC draft Branch Technical Position on concentration averaging of greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Tuite, P.; Tuite, K.; Harris, G. [Waste Management Group, Inc., Peekskill, NY (United States)

    1994-09-01T23:59:59.000Z

    This report evaluates the effects of concentration averaging practices on the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) generated by the nuclear utility industry and sealed sources. Using estimates of the number of waste components that individually exceed Class C limits, this report calculates the proportion that would be classified as GTCC LLW after applying concentration averaging; this proportion is called the concentration averaging factor. The report uses the guidance outlined in the 1993 Nuclear Regulatory Commission (NRC) draft Branch Technical Position on concentration averaging, as well as waste disposal experience at nuclear utilities, to calculate the concentration averaging factors for nuclear utility wastes. The report uses the 1993 NRC draft Branch Technical Position and the criteria from the Barnwell, South Carolina, LLW disposal site to calculate concentration averaging factors for sealed sources. The report addresses three waste groups: activated metals from light water reactors, process wastes from light-water reactors, and sealed sources. For each waste group, three concentration averaging cases are considered: high, base, and low. The base case, which is the most likely case to occur, assumes using the specific guidance given in the 1993 NRC draft Branch Technical Position on concentration averaging. To project future GTCC LLW generation, each waste category is assigned a concentration averaging factor for the high, base, and low cases.

  9. Crystallization during processing of nuclear waste glass

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2010-12-01T23:59:59.000Z

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glassmaking are reviewed.

  10. Memo - Legacy Technical Area (TA)-55 Nitrate Salt Wastes at TA...

    Office of Environmental Management (EM)

    Memo - Legacy Technical Area (TA)-55 Nitrate Salt Wastes at TA-54, Potential Applicability of Resource Conservation and Recovery Act (RCRA) D001D002D003 Waste Codes Memo - Legacy...

  11. Hanford site solid waste management environmental impact statement technical information document [SEC 1 THRU 4

    SciTech Connect (OSTI)

    FRITZ, L.L.

    2003-04-01T23:59:59.000Z

    This Technical Information Document (TID) provides engineering data to support DOE/EIS-0286, ''Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement,'' including assumptions and waste volumes calculation data.

  12. Mixing of incompatible materials in waste tanks technical basis document

    SciTech Connect (OSTI)

    SANDGREN, K.R.

    2003-03-21T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  13. The Very Deep Hole Concept: Evaluation of an Alternative for Nuclear Waste Disposal

    E-Print Network [OSTI]

    1979-01-01T23:59:59.000Z

    OF AN ALTERNATIVE FOR NUCLEAR WASTE DISPOSAL M.T. O'Brien,OF AN ALTERNATIVE FOR NUCLEAR WASTE DISPOSAL M. T. O'Brien,from commercial nuclear wastes in geologic storage. Oak

  14. Geothermal reservoir simulation to enhance confidence in predictions for nuclear waste disposal

    E-Print Network [OSTI]

    Kneafsey, Timothy J.; Pruess, Karsten; O'Sullivan, Michael J.; Bodvarsson, Gudmundur S.

    2002-01-01T23:59:59.000Z

    for high-level nuclear waste. Journal of Contaminantfor a Potential High-Level Nuclear Waste Repository at YuccaHeat Flow Near High-Level Nuclear Waste Packages Emplaced in

  15. REGIONAL THERMOHYDROLOGICAL EFFECTS OF AN UNDERGROUND REPOSITORY FOR NUCLEAR WASTES IN HARD ROCK

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2014-01-01T23:59:59.000Z

    underground repository for nuclear waste in hard rock, LBL-and Vath, J.E. , Nuclear waste projections and source-termthe Scientific Basis for Nuclear Waste Management, Material

  16. CRAD, New Nuclear Facility Documented Safety Analysis and Technical...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Technical Safety Requirements - December 2, 2014 (EA CRAD 31-07, Rev. 0) More Documents & Publications CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08...

  17. An Underwater Robotic Network for Monitoring Nuclear Waste Storage Pools

    E-Print Network [OSTI]

    Jeavons, Peter

    , it is still necessary to store this waste in cool- ing ponds for 20 to 60 years to remove the heatAn Underwater Robotic Network for Monitoring Nuclear Waste Storage Pools Sarfraz Nawaz1 , Muzammil spread with grow- ing world population. However, the radioactive waste generated in these power plants

  18. Review of Nuclear Safety Culture at the Hanford Site Waste Treatment...

    Office of Environmental Management (EM)

    Review of Nuclear Safety Culture at the Hanford Site Waste Treatment and Immobilization Plant Project, October 2010 Review of Nuclear Safety Culture at the Hanford Site Waste...

  19. Nuclear waste management. Quarterly progress report, April-June 1981

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A.

    1981-09-01T23:59:59.000Z

    Reports and summaries are presented for the following: high-level waste process development; alternative waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; and analysis of spent fuel policy implementation.

  20. Material Not Categorized As Waste (MNCAW) data report. Radioactive Waste Technical Support Program

    SciTech Connect (OSTI)

    Casey, C.; Heath, B.A.

    1992-11-01T23:59:59.000Z

    The Department of Energy (DOE), Headquarters, requested all DOE sites storing valuable materials to complete a questionnaire about each material that, if discarded, could be liable to regulation. The Radioactive Waste Technical Support Program entered completed questionnaires into a database and analyzed them for quantities and type of materials stored. This report discusses the data that TSP gathered. The report also discusses problems revealed by the questionnaires and future uses of the data. Appendices contain selected data about material reported.

  1. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    SciTech Connect (OSTI)

    Jooho, W.; Baldwin, G. T.

    2005-04-01T23:59:59.000Z

    One critical aspect of any denuclearization of the Democratic People’s Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for “complete, verifiable and irreversible dismantlement,” or “CVID.” It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long completion times. The radioactive waste management problem in fact offers a prospect for international participation to engage the DPRK constructively. DPRK nuclear dismantlement, when accompanied with a concerted effort for effective radioactive waste management, can be a mutually beneficial goal.

  2. Final Technical Report; NUCLEAR ENGINEERING RECRUITMENT EFFORT

    SciTech Connect (OSTI)

    Kerrick, Sharon S.; Vincent, Charles D.

    2007-07-02T23:59:59.000Z

    This report provides the summary of a project whose purpose was to support the costs of developing a nuclear engineering awareness program, an instruction program for teachers to integrate lessons on nuclear science and technology into their existing curricula, and web sites for the exchange of nuclear engineering career information and classroom materials. The specific objectives of the program were as follows: OBJECTIVE 1: INCREASE AWARENESS AND INTEREST OF NUCLEAR ENGINEERING; OBJECTIVE 2: INSTRUCT TEACHERS ON NUCLEAR TOPICS; OBJECTIVE 3: NUCLEAR EDUCATION PROGRAMS WEB-SITE; OBJECTIVE 4: SUPPORT TO UNIVERSITY/INDUSTRY MATCHING GRANTS AND REACTOR SHARING; OBJECTIVE 5: PILOT PROJECT; OBJECTIVE 6: NUCLEAR ENGINEERING ENROLLMENT SURVEY AT UNIVERSITIES

  3. Method of preparing nuclear wastes for tansportation and interim storage

    DOE Patents [OSTI]

    Bandyopadhyay, Gautam (Naperville, IL); Galvin, Thomas M. (Darien, IL)

    1984-01-01T23:59:59.000Z

    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.

  4. Nuclear waste vitrification efficiency: cold cap reactions

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Kruger, Albert A.; Pokorny, Richard

    2012-12-15T23:59:59.000Z

    The cost and schedule of nuclear waste treatment and immobilization are greatly affected by the rate of glass production. Various factors influence the performance of a waste-glass melter. One of the most significant, and also one of the least understood, is the process of batch melting. Studies are being conducted to gain fundamental understanding of the batch reactions, particularly those that influence the rate of melting, and models are being developed to link batch makeup and melter operation to the melting rate. Batch melting takes place within the cold cap, i.e., a batch layer floating on the surface of molten glass. The conversion of batch to glass consists of various chemical reactions, phase transitions, and diffusion-controlled processes. These include water evaporation (slurry feed contains as high as 60% water), gas evolution, the melting of salts, the formation of borate melt, reactions of borate melt with molten salts and with amorphous oxides (Fe2O3 and Al2O3), the formation of intermediate crystalline phases, the formation of a continuous glass-forming melt, the growth and collapse of primary foam, and the dissolution of residual solids. To this list we also need to add the formation of secondary foam that originates from molten glass but accumulates on the bottom of the cold cap. This study presents relevant data obtained for a high-level-waste melter feed and introduces a one-dimensional (1D) mathematical model of the cold cap as a step toward an advanced three-dimensional (3D) version for a complete model of the waste glass melter. The 1D model describes the batch-to-glass conversion within the cold cap as it progresses in a vertical direction. With constitutive equations and key parameters based on measured data, and simplified boundary conditions on the cold-cap interfaces with the glass melt and the plenum space of the melter, the model provides sensitivity analysis of the response of the cold cap to the batch makeup and melter conditions. The model demonstrates that batch foaming has a decisive influence on the rate of melting. Understanding the dynamics of the foam layer at the bottom of the cold cap and the heat transfer through it appears crucial for a reliable prediction of the rate of melting as a function of the melter-feed makeup and melter operation parameters. Although the study is focused on a batch for waste vitrification, the authors expect that the outcome will also be relevant for commercial glass melting.

  5. NUCLEAR WASTE VITRIFICATION EFFICIENCY COLD CAP REACTIONS

    SciTech Connect (OSTI)

    KRUGER AA; HRMA PR; POKORNY R

    2011-07-29T23:59:59.000Z

    The cost and schedule of nuclear waste treatment and immobilization are greatly affected by the rate of glass production. Various factors influence the performance of a waste-glass melter. One of the most significant, and also one of the least understood, is the process of batch melting. Studies are being conducted to gain fundamental understanding of the batch reactions, particularly those that influence the rate of melting, and models are being developed to link batch makeup and melter operation to the melting rate. Batch melting takes place within the cold cap, i.e., a batch layer floating on the surface of molten glass. The conversion of batch to glass consists of various chemical reactions, phase transitions, and diffusion-controlled processes. These include water evaporation (slurry feed contains as high as 60% water), gas evolution, the melting of salts, the formation of borate melt, reactions of borate melt with molten salts and with amorphous oxides (Fe{sub 2}O{sub 3} and Al{sub 2}O{sub 3}), the formation of intermediate crystalline phases, the formation of a continuous glass-forming melt, the growth and collapse of primary foam, and the dissolution of residual solids. To this list we also need to add the formation of secondary foam that originates from molten glass but accumulates on the bottom of the cold cap. This study presents relevant data obtained for a high-level-waste melter feed and introduces a one-dimensional (1D) mathematical model of the cold cap as a step toward an advanced three-dimensional (3D) version for a complete model of the waste glass melter. The 1D model describes the batch-to-glass conversion within the cold cap as it progresses in a vertical direction. With constitutive equations and key parameters based on measured data, and simplified boundary conditions on the cold-cap interfaces with the glass melt and the plenum space of the melter, the model provides sensitivity analysis of the response of the cold cap to the batch makeup and melter conditions. The model demonstrates that batch foaming has a decisive influence on the rate of melting. Understanding the dynamics of the foam layer at the bottom of the cold cap and the heat transfer through it appears crucial for a reliable prediction of the rate of melting as a function of the melter-feed makeup and melter operation parameters. Although the study is focused on a batch for waste vitrification, the authors expect that the outcome will also be relevant for commercial glass melting.

  6. Nuclear waste management. Quarterly progress report, April-June 1980

    SciTech Connect (OSTI)

    Platt, A.M.; Powell, J.A. (comps.)

    1980-09-01T23:59:59.000Z

    The status of the following programs is reported: high-level waste immobilization; alternative waste forms; Nuclear Waste Materials Characterization Center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of fission products in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; systems study on engineered barriers; criteria for defining waste isolation; spent fuel and fuel pool component integrity program; analysis of spent fuel policy implementation; asphalt emulsion sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and development of backfill material.

  7. Nuclear waste management. Quarterly progress report, October through December 1980

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1981-03-01T23:59:59.000Z

    Progress reports and summaries are presented under the following headings: high-level waste process development; alternative waste forms; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of radionuclides in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; high level waste form preparation; development of backfill material; development of structural engineered barriers; ONWI disposal charge analysis; spent fuel and fuel component integrity program; analysis of spent fuel policy implementation; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; revegetation of inactive uranium tailing sites; verification instrument development.

  8. Nuclear waste treatment program. Annual report for FY 1985

    SciTech Connect (OSTI)

    Powell, J.A. (ed.)

    1986-04-01T23:59:59.000Z

    Two of the US Department of Energy's (DOE) nuclear waste management-related goals are: (1) to ensure that waste management is not an obstacle to the further deployment of light-water reactors (LWR) and the closure of the nuclear fuel cycle and (2) to fulfill its institutional responsibility for providing safe storage and disposal of existing and future nuclear wastes. As part of its approach to achieving these goals, the Office of Terminal Waste Disposal and Remedial Action of DOE established what is now called the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL) during the second half of FY 1982. To support DOE's attainment of its goals, the NWTP is to provide (1) documented technology necessary for the design and operation of nuclear waste treatment facilities by commercial enterprises as part of a licensed waste management system and (2) problem-specific treatment approaches, waste form and treatment process adaptations, equipment designs, and trouble-shooting assistance, as required, to treat existing wastes. This annual report describes progress during FY 1985 toward meeting these two objectives. The detailed presentation is organized according to the task structure of the program.

  9. Nuclear Waste Management Program summary document, FY 1981

    SciTech Connect (OSTI)

    Meyers, Sheldon

    1980-03-01T23:59:59.000Z

    The Nuclear Waste Management Program Summary Document outlines the operational and research and development (R and D) activities of the Office of Nuclear Waste Management (NEW) under the Assistant Secretary for Nuclear Energy, US Department of Energy (DOE). This document focuses on the current and planned activities in waste management for FY 1981. This Program Summary Document (PSD) was prepared in order to explain the Federal nuclear waste management and spent fuel storage programs to Congress and its committees and to interested members of the public, the private sector, and the research community. The national energy policy as it applies to waste management and spent fuel storage is presented first. The program strategy, structure, budget, management approach, and public participation programs are then identified. The next section describes program activities and outlines their status. Finally, the applicability of departmental policies to NEW programs is summarized, including field and regional activities, commercialization plans, and environmental and socioeconomic implications of waste management activities, and international programs. This Nuclear Waste Management Program Summary Document is meant to serve as a guide to the progress of R and D and other energy technology programs in radioactive waste management. The R and D objective is to provide the Nation with acceptable solutions to short- and long-term management problems for all forms of radioactive waste and spent fuel.

  10. Technical Safety Requirements for the B695 Segment of the Decontamination and Waste Treatment Facility

    SciTech Connect (OSTI)

    Larson, H L

    2007-09-07T23:59:59.000Z

    This document contains Technical Safety Requirements (TSRs) for the Radioactive and Hazardous Waste Management (RHWM) Division's B695 Segment of the Decontamination and Waste Treatment Facility (DWTF) at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the B695 Segment of the DWTF. The TSRs are derived from the Documented Safety Analysis (DSA) for the B695 Segment of the DWTF (LLNL 2004). The analysis presented there determined that the B695 Segment of the DWTF is a low-chemical hazard, Hazard Category 3, nonreactor nuclear facility. The TSRs consist primarily of inventory limits as well as controls to preserve the underlying assumptions in the hazard analyses. Furthermore, appropriate commitments to safety programs are presented in the administrative controls section of the TSRs. The B695 Segment of the DWTF (B695 and the west portion of B696) is a waste treatment and storage facility located in the northeast quadrant of the LLNL main site. The approximate area and boundary of the B695 Segment of the DWTF are shown in the B695 Segment of the DWTF DSA. Activities typically conducted in the B695 Segment of the DWTF include container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. B695 is used to store and treat radioactive, mixed, and hazardous waste, and it also contains equipment used in conjunction with waste processing operations to treat various liquid and solid wastes. The portion of the building called Building 696 Solid Waste Processing Area (SWPA), also referred to as B696S in this report, is used primarily to manage solid radioactive waste. Operations specific to the SWPA include sorting and segregating low-level waste (LLW) and transuranic (TRU) waste, lab-packing, sampling, and crushing empty drums that previously contained LLW. A permit modification for B696S was submitted to DTSC in January 2004 to store and treat hazardous and mixed waste. Upon approval of the permit modification, B696S rooms 1007, 1008, and 1009 will be able to store hazardous and mixed waste for up to 1 year. Furthermore, an additional drum crusher and a Waste Packaging Unit will be permitted to treat hazardous and mixed waste. RHWM generally processes LLW with no, or extremely low, concentrations of transuranics (i.e., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, {sup 3}H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment of the DWTF. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment of the DWTF. A basis explanation follows each of the requirements described in Section 5.5, Specific Administrative Controls. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6.

  11. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03T23:59:59.000Z

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  12. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1998-01-01T23:59:59.000Z

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  13. acidic nuclear wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    acidic nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear ""BurningBurning""...

  14. automated nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    automated nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear...

  15. acidic nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    acidic nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear ""BurningBurning""...

  16. Improved Management of the Technical Interfaces Between the Hanford Tank Farm Operator and the Hanford Waste Treatment Plant - 13383

    SciTech Connect (OSTI)

    Duncan, Garth M. [Bechtel National Inc., 2435 Stevens Center Place, Richland, Washington, 99352 (United States)] [Bechtel National Inc., 2435 Stevens Center Place, Richland, Washington, 99352 (United States); Saunders, Scott A. [Washington River Protection Solutions, P.O. Box 850, Richland, Washington, 99352 (United States)] [Washington River Protection Solutions, P.O. Box 850, Richland, Washington, 99352 (United States)

    2013-07-01T23:59:59.000Z

    The Department of Energy (DOE) is constructing the Waste Treatment and Immobilization Plant (WTP) at the Hanford site in Washington to treat and immobilize approximately 114 million gallons of high level radioactive waste (after all retrievals are accomplished). In order for the WTP to be designed and operated successfully, close coordination between the WTP engineering, procurement, and construction contractor, Bechtel National, Inc. and the tank farms operating contractor (TOC), Washington River Protection Solutions, LLC, is necessary. To develop optimal solutions for DOE and for the treatment of the waste, it is important to deal with the fact that two different prime contractors, with somewhat differing contracts, are tasked with retrieving and delivering the waste and for treating and immobilizing that waste. The WTP and the TOC have over the years cooperated to manage the technical interface. To manage what is becoming a much more complicated interface as the WTP design progresses and new technical issues have been identified, an organizational change was made by WTP and TOC in November of 2011. This organizational change created a co-located integrated project team (IPT) to deal with mutual and interface issues. The Technical Organization within the One System IPT includes employees from both TOC and WTP. This team has worked on a variety of technical issues of mutual interest and concern. Technical issues currently being addressed include: - The waste acceptance criteria; - Waste feed delivery and the associated data quality objectives (DQO); - Evaluation of the effects of performing a riser cut on a single shell tank on WTP operations; - The disposition of secondary waste from both TOC and WTP; - The close coordination of the TOC double shell tank mixing and sampling program and the Large Scale Integrated Test (LSIT) program for pulse jet mixers at WTP along with the associated responses to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2; - Development of a set of alternatives to the current baseline that involve aspects of direct feed, feed conditioning, and design changes. The One System Technical Organization has served WTP, TOC, and DOE well in managing and resolving issues at the interface. This paper describes the organizational structure used to improve the interface and several examples of technical interface issues that have been successfully addressed by the new organization. (authors)

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

    SciTech Connect (OSTI)

    Dees, L.A.

    1994-08-15T23:59:59.000Z

    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.

  18. DRAFT TECHNICAL GUIDANCE DOCUMENT ON STATIC AND SEISMIC SLOPE STABILITY FOR SOLID WASTE

    E-Print Network [OSTI]

    DRAFT TECHNICAL GUIDANCE DOCUMENT ON STATIC AND SEISMIC SLOPE STABILITY FOR SOLID WASTE CONTAINMENT FACILITIES PRODUCED BY THE SOLID WASTE MANAGEMENT PROGRAM DIVISION OF ENVIRONMENTAL QUALITY MISSOURI IN GEOSYNTHETIC MATERIALS Page 33 5.0 ENGINEERING PROPERTIES OF MUNICIPAL SOLID WASTE Page 36 I. STATIC PROPERTIES

  19. Design of the Prototypical Cryomodule for the EUROTRANS Superconducting Linac for Nuclear Waste Transmutation

    E-Print Network [OSTI]

    Barbanotti, S; Blache, P; Commeaux, C; Duthil, P; Panzeri, N; Pierini, P; Rampnoux, E; Souli, M

    2008-01-01T23:59:59.000Z

    Design of the Prototypical Cryomodule for the EUROTRANS Superconducting Linac for Nuclear Waste Transmutation

  20. AISI waste oxide recycling program. Final technical report

    SciTech Connect (OSTI)

    Aukrust, E.; Downing, K.B.; Sarma, B.

    1995-08-01T23:59:59.000Z

    In March 1995 AISI completed a five-year, $60 million collaborative development program on Direct Steelmaking cost-shared by DOE under the Metals Initiative. This program defined an energy-efficient and environmentally-friendly technology to produce hot metal for steelmaking directly from coal and iron ore pellets without incurring the high capital costs and environmental problems associated with traditional coke oven and blast furnace technology. As it becomes necessary to replace present capacity, this new technology will be favored because of reduced capital costs, higher energy efficiency, and lower operating costs. In April 1994, having failed to move forward with a demonstration plant for direct ironmaking, despite substantial efforts by both Stelco and Geneva Steel, an alternative opportunity was sought to commercialize this new technology without waiting until existing ironmaking capacity needed to be replaced. Recycling and resource recovery of steel plant waste oxides was considered an attractive possibility. This led to approval of a ten-month, $8.3 million joint program with DOE on recycling steel plant waste oxides utilizing this new smelting technology. This highly successful trial program was completed in December 1994. The results of the pilot plant work and a feasibility study for a recycling demonstration plant are presented in this final technical report.

  1. Canister design for deep borehole disposal of nuclear waste

    E-Print Network [OSTI]

    Hoag, Christopher Ian

    2006-01-01T23:59:59.000Z

    The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas, and geothermal drilling ...

  2. ANL technical support program for DOE Environmental Restoration and Waste Management. Annual report, October 1991--September 1992

    SciTech Connect (OSTI)

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J. [Argonne National Lab., IL (United States); Bourcier, W.L.; Morgan, L.E.; Newton, L.; Nielsen, J.K.; Phillips, B.L. [Lawrence Livermore National Lab., CA (United States); Ewing, R.C.; Wang, L. [Univ. of New Mexico, Albuquerque, NM (United States); Li, H.; Tomozawa, M. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    1993-05-01T23:59:59.000Z

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1992 on the following tasks: 1. A compendium of the characteristics of high-level nuclear waste borosilicate glass has been written. 2. A critical review of important parameters that affect the reactivity of glass in an unsaturated environment is being prepared. 3. A series of tests has been started to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. 4. The effect of radiation upon the durability of waste glasses at a high glass surface area-to-liquid volume (SA/V) ratio and a high gas-to-liquid volume ratio will be assessed. These tests address both vapor and high SA/V liquid conditions. 5. A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SAN ratios. Such differences in the SAN ratio may significantly affect glass durability. 6. A series of natural analogue tests is being analyzed to demonstrate a meaningful relationship between experimental and natural alteration conditions. 7. Analytical electron microscopy (AEM), infrared spectroscopys and nuclear resonant profiling are being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Additionally, colloids from the leach solutions are being studied using AEM. 8. A technical review of AEM results is being provided. 9. A study of water diffusion involving nuclear waste glasses is being performed. 10. A mechanistically based model is being developed to predict the performance of glass over repository-relevant time periods.

  3. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, Terry R. (Wheaton, IL); Ackerman, John P. (Downers Grove, IL); Tomczuk, Zygmunt (Orland Park, IL); Fischer, Donald F. (Glen Ellyn, IL)

    1989-01-01T23:59:59.000Z

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

  4. Lessons in waste minimization from nuclear industry experience

    SciTech Connect (OSTI)

    Devgun, J.S.; Thuot, J.R.; Vrtis, J.

    1996-07-01T23:59:59.000Z

    The nuclear power industry has been very successful at reducing waste volumes and waste sources. The success has been driven by escalating cost, decreasing disposal ability, and a desire by the industry to achieve excellence. The result has been a cycle of continuing improvement resulting in reduced cost. Many of the examples of Dry Active Waste reduction are applicable to the Department of Energy in both operations and remedial activities. This paper discusses several successful examples of utility applications in this area.

  5. Method for forming microspheres for encapsulation of nuclear waste

    DOE Patents [OSTI]

    Angelini, Peter (Oak Ridge, TN); Caputo, Anthony J. (Knoxville, TN); Hutchens, Richard E. (Knoxville, TN); Lackey, Walter J. (Oak Ridge, TN); Stinton, David P. (Knoxville, TN)

    1984-01-01T23:59:59.000Z

    Microspheres for nuclear waste storage are formed by gelling droplets containing the waste in a gelation fluid, transferring the gelled droplets to a furnace without the washing step previously used, and heating the unwashed gelled droplets in the furnace under temperature or humidity conditions that result in a substantially linear rate of removal of volatile components therefrom.

  6. National Nuclear Security Administration Babcock & Wilcox Technical

    National Nuclear Security Administration (NNSA)

    France, and Italy. B&W Y-12 led efforts with TVA, AREVA, Nuclear Fuel Services, and the Savannah River Site (SRS) to coordinate the resumption of the Blended Low- Enriched...

  7. National Nuclear Security Administration Babcock & Wilcox Technical

    National Nuclear Security Administration (NNSA)

    which will power more than 60% of the Pantex Plant in the future with clean, renewable wind energy. B&W Pantex managed security challenges, protecting special nuclear material,...

  8. Water borne transport of high level nuclear waste in very deep borehole disposal of high level nuclear waste

    E-Print Network [OSTI]

    Cabeche, Dion Tunick

    2011-01-01T23:59:59.000Z

    The purpose of this report is to examine the feasibility of the very deep borehole experiment and to determine if it is a reasonable method of storing high level nuclear waste for an extended period of time. The objective ...

  9. Waste Management Program. Technical progress report, October-December 1982

    SciTech Connect (OSTI)

    None

    1983-07-01T23:59:59.000Z

    This quarterly report provides current information on operations and development programs for the management of radioactive wastes from operation of the Savannah River Plant and offplant participants. The studies on environmental and safety assessments, in situ storage or disposal, waste from development and characterization, process and equipment development, and low-level waste management are a part of the Long-Term Waste Management Technology Program. The following studies are reported for the SR Interim Waste Operations Program: surveillance and maintenance, waste concentration, low-level effluent waste, tank replacement/waste transfer, and solid waste storage and related activities.

  10. Waste Management Program. Technical progress report, Aporil-June 1983

    SciTech Connect (OSTI)

    None

    1984-02-01T23:59:59.000Z

    This quarterly report provides current information on operations and development programs for the management of radioactive wastes from operation of the Savannah River Plant. The studies on environmental and safety assessments, process and equipment development, TRU waste, and low-level waste are a part of the Long-Term Waste Management Technology Program. The following studies are reported for the SR Interim Waste Operations Program: surveillance and maintenance, waste concentration, low-level effluent waste, tank replacement/waste transfer, and solid waste storage and related activities.

  11. Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  12. US Department of Energy mixed waste characterization, treatment, and disposal focus area technical baseline development process

    SciTech Connect (OSTI)

    Roach, J.A.; Gombert, D. [Lockheed Martin Idaho Technologies, Idaho Falls, ID (United States)

    1996-12-31T23:59:59.000Z

    The US Department of Energy (DOE) created the Mixed Waste Characterization, Treatment, and Disposal Focus Area (MWFA) to develop and facilitate implementation of technologies required to meet its commitments for treatment of mixed wastes under the Federal Facility Compliance Act (FFCA), and in accordance with the Land Disposal Restrictions (LDR) of the Resource Conservation and Recovery Act (RCRA). Mixed wastes include both mixed low-level waste (MLLW) and mixed transuranic (MTRU) waste. The goal of the MWFA is to develop mixed waste treatment systems to the point of implementation by the Environmental Management (EM) customer. To accomplish this goal, the MWFA is utilizing a three step process. First, the treatment system technology deficiencies were identified and categorized. Second, these identified needs were prioritized. This resulted in a list of technical deficiencies that will be used to develop a technical baseline. The third step, the Technical Baseline Development Process, is currently ongoing. When finalized, the technical baseline will integrate the requirements associated with the identified needs into the planned and ongoing environmental research and technology development activities supported by the MWFA. Completion of this three-step process will result in a comprehensive technology development program that addresses customer identified and prioritized needs. The MWFA technical baseline will be a cost-effective, technically-defensible tool for addressing and resolving DOE`s mixed waste problems.

  13. alumina nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    alumina nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Nuclear magnetic resonance-based...

  14. NuclearNuclear ""BurningBurning"" of Nuclearof Nuclear ""WasteWaste"" Constantine P. Tzanos

    E-Print Network [OSTI]

    , total of about 66 kilometers long, to accommodate about 1100 waste packages, 70,000 tons of heavy metal 300 meters above water table. s Construction: 5 years s Operations: 50 years s Monitoring: 50 years failure, igneousigneous intrusion, volcanic eruption, seismic ground motion, and seismicintrusion

  15. Nuclear physics information needed for accelerator driven transmutation of nuclear waste

    SciTech Connect (OSTI)

    Lisowski, P.W.; Bowman, C.D.; Arthur, E.D.; Young, P.G.

    1991-01-01T23:59:59.000Z

    There is renewed interest in using accelerator driven neutron sources to address the problem of high-level long-lived nuclear waste. Several laboratories have developed systems that may have a significant impact on the future use of nuclear power, adding options for dealing with long-lived actinide wastes and fission products, and for power production. This paper describes a new Los Alamos concept using thermal neutrons and examines the nuclear data requirements. 7 refs., 3 figs., 1 tab.

  16. Conversion of transuranic waste to low level waste by decontamination: a technical and economic evaluation

    SciTech Connect (OSTI)

    Allen, R.P.; Hazelton, R.F.

    1984-12-01T23:59:59.000Z

    A study was conducted to evaluate the technical and economic feasibility of using in-situ decontamination techniques to convert glove boxes and other large TRU-contaminated components directly into LLW. The results of the technical evaluation indicate that in-situ decontamination of these types of components to non-TRU levels is technically feasible. Applicable decontamination techniques include electropolishing, hand scrubbing, chemical washes/sprays, strippable coatings and Freon spray-cleaning. The removal of contamination from crevices and other holdup areas remains a problem, but may be solved through further advances in decontamination technology. Also, the increase in the allowable maximum TRU level from 10 nCi/g to 100 nCi/g as defined in DOE Order 5820.2 reduces the removal requirement and facilitates measurement of the remaining quantities. The major emphasis of the study was on a cost/benefit evaluation that included a review and update of previous analyses and evaluations of TRU-waste volume reduction and conversion options. The results of the economic evaluation show, for the assumptions used, that there is a definite cost incentive to size reduce large components, and that decontamination of sectioned material has become cost competitive with the size reduction options. In-situ decontamination appears to be the lowest cost option when based on routine-type operations conducted by well-trained and properly equipped personnel. 16 references, 1 figure, 7 tables.

  17. Nuclear waste policy and public acceptance in France

    SciTech Connect (OSTI)

    Guais, J.C. [NUSYS, Paris (France)

    1993-12-31T23:59:59.000Z

    In France, the development of an extensive nuclear program has traditionally met relative support from the public and the politicals. Yet, facing some unusual opposition in 1991, the government declared a moratorium on the selection process of a geological repository. The authors review in this paper the successive steps of the nuclear waste storage program over the last 12 years, from the successful siting of two LLW storage facilities by ANDRA, the national agency for the storage of nuclear waste, to the more difficult years of the search for a suitable site to host the HLW repository which led to a new approach of the issue.

  18. Progress and Status of the Ignalina Nuclear Power Plant's New Solid Waste Management and Storage Facilities

    SciTech Connect (OSTI)

    Rausch, J.; Henderson, R.W. [NUKEM Technologies GmbH, Alzenau (Germany); Penkov, V. [State Enterprise Ignalina Nuclear Power Plant, Visaginas (Lithuania)

    2008-07-01T23:59:59.000Z

    A considerable amount of dry radioactive waste from former NPP operation has accumulated up to date and is presently stored at the Ignalina NPP site, Lithuania. Current storage capacities are nearly exhausted and more waste is to come from future decommissioning of the two RMBKtype reactors. Additionally, the existing storage facilities does not comply to the state-of-the-art technology for handling and storage of radioactive waste. In 2005, INPP faced this situation of a need for waste processing and subsequent interim storage of these wastes by contracting NUKEM with the design, construction, installation and commissioning of new waste management and storage facilities. The subject of this paper is to describe the scope and the status of the new solid waste management and storage facilities at the Ignalina Nuclear Power Plant. In summary: The turnkey contract for the design, supply and commission of the SWMSF was awarded in December 2005. The realisation of the project was initially planned within 48 month. The basic design was finished in August 2007 and the Technical Design Documentation and Preliminary Safety Analyses Report was provided to Authorities in October 2007. The construction license is expected in July 2008. The procurement phase was started in August 2007, start of onsite activities is expected in November 2007. The start of operation of the SWMSF is scheduled for end of 2009. (authors)

  19. Thermally induced mechanical and permeability changes around a nuclear waste repository -- a far-field study based on equivalent properties determined by a discrete approach

    E-Print Network [OSTI]

    Min, Ki-Bok; Rutqvist, Jonny; Tsang, Chin-Fu; Jing, Lanru

    2004-01-01T23:59:59.000Z

    changes around a nuclear waste repository – a far-fieldmass containing a hypothetical nuclear waste repository. Thethe safe isolation of nuclear wastes from the biosphere,

  20. Bridging the Gap in the Chemical Thermodynamic Database for Nuclear Waste Repository: Studies of the Effect of Temperature on Actinide Complexation

    E-Print Network [OSTI]

    Rao, Linfeng

    2011-01-01T23:59:59.000Z

    Thermodynamic Database for Nuclear Waste Repository: Studiesthermodynamic database for nuclear waste repository wherethe safe management of nuclear wastes is to store the high-

  1. Multiple-code benchmark simulation study of coupled THMC processes in the excavation disturbed zone associated with geological nuclear waste repositories

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    filled and open-drift nuclear waste repositories in Task DASSOCIATED WITH GEOLOGICAL NUCLEAR WASTE REPOSITORIES J.emplacement drifts of a nuclear waste repository. This BMT

  2. Deployment at the Savannah River Site of a standardized, modular transportable and connectable hazard category 2 nuclear system for repackaging TRU waste

    SciTech Connect (OSTI)

    Lussiez, G. (Guy); Hickman, S. (Scott); Anast, K. R. (Kurt R.); Oliver, W. B. (William B.)

    2004-01-01T23:59:59.000Z

    This paper describes the conception, design, fabrication and deployment of a modular, transportable, connectable Category 2 nuclear system deployed at the Savannah River site to be used for characterizing and repackaging Transuranic Waste destined for the Waste Isolation Pilot Plant (WIPP). A standardized Nuclear Category 2 and Performance Category 2 envelope called a 'Nuclear Transportainer' was conceived and designed that provides a safety envelope for nuclear operations. The Nuclear Transportainer can be outfitted with equipment that performs functions necessary to meet mission objectives, in this case repackaging waste for shipment to WIPP. Once outfitted with process and ventilation systems the Nuclear Transportainer is a Modular Unit (MU). Each MU is connectable to other MUS - nuclear or non-nuclear - allowing for multiple functions, command & control, or increasing capacity. The design took advantage of work already in-progress at Los Alamos National Laboratory (LANL) for a similar system to be deployed at LANL's Technical Area 54.

  3. Nuclear Waste Disposal: An Independent View of the Big Picture and a Proposal for CARD

    E-Print Network [OSTI]

    California at Santa Cruz, University of

    1 Nuclear Waste Disposal: An Independent View of the Big Picture and a Proposal for CARD Presented to isolate nuclear waste successfully from the biosphere for the long term can be developed if our society to this impasse? In the 1940's at the beginning of the nuclear age, nuclear waste was seen as a "problem" only

  4. An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions

    E-Print Network [OSTI]

    Sheffield, University of

    An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been originating from the reprocessing of spent nuclear fuel. Experiments using simulated nuclear waste streams

  5. ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL

    E-Print Network [OSTI]

    ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL and Waste Management Co.) for encapsulation of nuclear waste. Due to the radiation emitted by the nuclear, and characterization. The applicability of linear array technique for inspection of copper lined canisters for nuclear

  6. Radioactive and chemotoxic wastes: Only radioactive wastes?

    SciTech Connect (OSTI)

    Eletti, G.F.; Tocci, M. [ENEA DISP, Rome (Italy)

    1993-12-31T23:59:59.000Z

    Radioactive waste arising from Italian Nuclear Power Plants and Research Centers, classified as 1st and 2nd Category wastes, are managed only as radioactive wastes following the Technical Guide No. 26 issued by the Italian Regulatory Body: ENEA DISP on 1987. A very important Regulatory Regime revision for Italian Nuclear Activities started at the end of 1991. This paper considers the need to develop a new strategy dedicated to mixed waste in line with current international trends.

  7. Underground nuclear energy complexes - technical and economic advantages

    SciTech Connect (OSTI)

    Myers, Carl W [Los Alamos National Laboratory; Kunze, Jay F [IDAHO STATE UNIV; Giraud, Kellen M [BABECOCK AND WILCOX; Mahar, James M [IDAHO STATE UNIV

    2010-01-01T23:59:59.000Z

    Underground nuclear power plant parks have been projected to be economically feasible compared to above ground instalIations. This paper includes a thorough cost analysis of the savings, compared to above ground facilities, resulting from in-place entombment (decommissioning) of facilities at the end of their life. reduced costs of security for the lifetime of the various facilities in the underground park. reduced transportation costs. and reduced costs in the operation of the waste storage complex (also underground). compared to the fair share of the costs of operating a national waste repository.

  8. Nevada Nuclear-Waste-Storage Investigations. Quarterly report, April-June 1982

    SciTech Connect (OSTI)

    None

    1982-09-01T23:59:59.000Z

    The Nevada Nuclear Waste Storage Investigations (NNWSI) are studying the Nevada Test Site (NTS) area to establish whether it would qualify as a licensable location for a commercial nuclear waste repository; determining whether specific underground rock masses in the NTS area are technically acceptable for permanently disposing of highly radioactive solid wastes; and developing and demonstrating the capability to safely handle and store commercial spent reactor fuel and high-level waste. Progress reports for the following eight tasks are presented: systems; waste package; site; repository; regulatory and institutional; test facilities; land acquisition; and program management. Some of the highlights are: A code library was established to provide a central location for documentation of repository performance assessment codes. A two-dimensional finite element code, SAGUARO, was developed for modeling saturated/unsaturated groundwater flow. The results of an initial experiment to determine canister penetration rates due to corrosion indicate the expected strong effect of toxic environmental conditions on the corrosion rate of carbon steel in tuff-conditioned water. Wells USW-H3 and USW-H4 at Yucca Mountain have been sampled for groundwater analysis. A summary characterizing and relating the mineralogy and petrology of Yucca Mountain tuffs was compiled from the findings of studies of core samples from five drill holes.

  9. Nuclear Energy Technical Assistance | Department of Energy

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /7 This is aLinksNuclear Energy

  10. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    E-Print Network [OSTI]

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01T23:59:59.000Z

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  11. NUCLEAR WASTE GLASSES CONTINUOUS MELTING AND BULK VITRIFICAITON

    SciTech Connect (OSTI)

    KRUGER AA; HRMA PR

    2008-03-24T23:59:59.000Z

    This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed.

  12. Nuclear Waste Glasses: Continuous Melting and Bulk Vitrification

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Kruger, Albert A.

    2008-02-25T23:59:59.000Z

    This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee that the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; on cooling, the glass may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed.

  13. Nuclear Waste Glasses: Continuous Melting and Bulk Vitrification

    SciTech Connect (OSTI)

    Hrma, Pavel R.; Kruger, Albert A.

    2009-01-15T23:59:59.000Z

    This contribution addresses various aspects of nuclear waste vitrification. Composition of nuclear wastes varies in the number of components and their composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee that the glass is easily made and resists environmental degradation. Glass formulation is facilitated by developing property-composition models. The strategy of model development and application is reviewed. However, the large variability of waste composition presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slows down the process; molten salts may accumulate in container refractory walls; on cooling, the glass may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed.

  14. National briefing summaries: Nuclear fuel cycle and waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

    1991-04-01T23:59:59.000Z

    Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

  15. Results from an International Simulation Study on Coupled Thermal, Hydrological, and Mechanical (THM) Processes near Geological Nuclear Waste Repositories

    E-Print Network [OSTI]

    Rutqvist, J.

    2008-01-01T23:59:59.000Z

    safety of a hypothetical nuclear waste repository – BMT1 ofAssociated with Nuclear Waste Repositories, Academic Press,safety of a hypothetical nuclear waste repository – BMT1 of

  16. A STUDY OF REGIONAL TEMPERATURE AND THERMOHYDROLOGICAL EFFECTS OF AN UNDERGROUND REPOSITORY FOR NUCLEAR WASTES IN HARD ROCK

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    generated by the stored nuclear wastes. Spent fuel 10-yeartask force for review of nuclear waste manage­ ment, DOE/ER-Review Group on Nuclear Waste Management. Subground report

  17. Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source

    E-Print Network [OSTI]

    Belanger, David P.

    1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source renewed interest amongst the nuclear science community as the debate over nuclear waste has increased .................................................................................27 2.1.2 Waste Minimization

  18. A STUDY OF REGIONAL TEMPERATURE AND THERMOHYDROLOGICAL EFFECTS OF AN UNDERGROUND REPOSITORY FOR NUCLEAR WASTES IN HARD ROCK

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    of heat generated by the stored nuclear wastes. Spent fuelmode of heat transfer from the nuclear waste to the rockdensity of heat generated by the stored nuclear wastes Fig.

  19. Nuclear waste management. Quarterly progress report, January-March, 1981

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A. (comp.)

    1981-06-01T23:59:59.000Z

    Reports and summaries are provided for the following programs: high-level waste process development; alternative waste forms; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclide in soils; low-level waste generation reduction handbook; waste management system studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

  20. Reducing nuclear danger through intergovernmental technical exchanges on nuclear materials safety management

    SciTech Connect (OSTI)

    Jardine, L.J. [Lawrence Livermore National Lab., CA (United States); Peddicord, K.L. [Texas A and M Univ., College Station, TX (United States); Witmer, F.E.; Krumpe, P.F. [USDOE, Washington, DC (United States); Lazarev, L.; Moshkov, M. [Radievyj Inst., Leningrad (Russian Federation)

    1997-04-09T23:59:59.000Z

    The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and Russian Minatom organizations.are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an improved and sustained common safety culture for handling these materials. An initiative that develops and uses personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment.

  1. Disposal of radioactive waste from nuclear research facilities

    E-Print Network [OSTI]

    Maxeiner, H; Kolbe, E

    2003-01-01T23:59:59.000Z

    Swiss radioactive wastes originate from nuclear power plants (NPP) and from medicine (e.g. radiation sources), industry (e.g. fire detectors) and research (e.g. CERN, PSI). Their conditioning, characterisation and documentation has to meet the demands given by the Swiss regulatory authorities including all information needed for a safe disposal in future repositories. For NPP wastes, arisings as well as the processes responsible for the buildup of short and long lived radionuclides are well known, and the conditioning procedures are established. The radiological inventories are determined on a routinely basis using a combined system of measurements and calculational programs. For waste from research, the situation is more complicated. The wide spectrum of different installations combined with a poorly known history of primary and secondary radiation results in heterogeneous waste sorts with radiological inventories quite different from NPP waste and difficult to measure long lived radionuclides. In order to c...

  2. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1988-07-12T23:59:59.000Z

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  3. Process to separate transuranic elements from nuclear waste

    DOE Patents [OSTI]

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1989-03-21T23:59:59.000Z

    A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  4. Scenarios for exercising technical approaches to verified nuclear reductions

    SciTech Connect (OSTI)

    Doyle, James [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    Presidents Obama and Medvedev in April 2009 committed to a continuing process of step-by-step nuclear arms reductions beyond the new START treaty that was signed April 8, 2010 and to the eventual goal of a world free of nuclear weapons. In addition, the US Nuclear Posture review released April 6, 2010 commits the US to initiate a comprehensive national research and development program to support continued progress toward a world free of nuclear weapons, including expanded work on verification technologies and the development of transparency measures. It is impossible to predict the specific directions that US-RU nuclear arms reductions will take over the 5-10 years. Additional bilateral treaties could be reached requiring effective verification as indicated by statements made by the Obama administration. There could also be transparency agreements or other initiatives (unilateral, bilateral or multilateral) that require monitoring with a standard of verification lower than formal arms control, but still needing to establish confidence to domestic, bilateral and multilateral audiences that declared actions are implemented. The US Nuclear Posture Review and other statements give some indication of the kinds of actions and declarations that may need to be confirmed in a bilateral or multilateral setting. Several new elements of the nuclear arsenals could be directly limited. For example, it is likely that both strategic and nonstrategic nuclear warheads (deployed and in storage), warhead components, and aggregate stocks of such items could be accountable under a future treaty or transparency agreement. In addition, new initiatives or agreements may require the verified dismantlement of a certain number of nuclear warheads over a specified time period. Eventually procedures for confirming the elimination of nuclear warheads, components and fissile materials from military stocks will need to be established. This paper is intended to provide useful background information for establishing a conceptual approach to a five-year technical program plan for research and development of nuclear arms reductions verification and transparency technologies and procedures.

  5. GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    in "Solid progress in nuclear waste treatment," NewTreatment, Conditioning and Storage of Solid Alpha-Bearing WasteTreatment, Conditioning and Storage of Solid-Alpha Bearing Waste

  6. DESIGN OF THE PROTOTYPICAL CRYOMODULE FOR THE EUROTRANS SUPERCONDUCTING LINAC FOR NUCLEAR WASTE

    E-Print Network [OSTI]

    Boyer, Edmond

    DESIGN OF THE PROTOTYPICAL CRYOMODULE FOR THE EUROTRANS SUPERCONDUCTING LINAC FOR NUCLEAR WASTE of the accelerator workpackage of the EUROTRANS program for the design of a nuclear waste transmutation system

  7. Joint Technical Operations Team | National Nuclear Security Administra...

    National Nuclear Security Administration (NNSA)

    Render Safe Joint Technical Operations Team Joint Technical Operations Team JTOT Logo NNSA's Joint Technical Operations Team (JTOT) provides specialized technical...

  8. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  9. Numerical methods for the simulation of a corrosion model in a nuclear waste deep repository $

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Numerical methods for the simulation of a corrosion model in a nuclear waste deep repository $ C of the French nuclear waste management agency ANDRA, investigations are conducted to optimize and finalize by the Nuclear Waste Management Agency ANDRA Corresponding author. Phone: +49 30 20372 560, Fax: +49 30 2044975

  10. Seismic modeling and analysis of a prototype heated nuclear waste storage tunnel, Yucca Mountain, Nevada

    E-Print Network [OSTI]

    Snieder, Roel

    Seismic modeling and analysis of a prototype heated nuclear waste storage tunnel, Yucca Mountain was heated to replicate the effects of long-term storage of decaying nuclear waste and to study the effects for the long- term storage of high-level nuclear waste from reactors and decom- missioned atomic weapons

  11. A ThreeDimensional Finite Element Simulation for Transport of Nuclear Waste Contamination in Porous Media

    E-Print Network [OSTI]

    Ewing, Richard E.

    A Three­Dimensional Finite Element Simulation for Transport of Nuclear Waste Contamination of South Carolina, Columbia, South Carolina ABSTRACT: Model equations for transport of nuclear­waste based up on the inherent physics. A three­dimensional finite element method for nuclear waste

  12. Reproductive Life Events in the Population Living in the Vicinity of a Nuclear Waste Reprocessing Plant

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Reproductive Life Events in the Population Living in the Vicinity of a Nuclear Waste Reprocessing: There is concern about the health of populations living close to nuclear waste reprocessing plants. We conducted a comparative study on reproductive life events in the general population living near the nuclear waste

  13. Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be

    E-Print Network [OSTI]

    Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be sited geological nuclear waste repository. There a suspicion of predetermination because the only district that has. National and international guidance on how best to select potential sites for deep geological nuclear waste

  14. SWAMI: An Autonomous Mobile Robot for Inspection of Nuclear Waste Storage Facilities

    E-Print Network [OSTI]

    Stephens, Larry M.

    SWAMI: An Autonomous Mobile Robot for Inspection of Nuclear Waste Storage Facilities Ron Fulbright Inspector (SWAMI) is a prototype mobile robot designed to perform autonomous inspection of nuclear waste user interface building tool called UIM/X. Introduction Safe disposal of nuclear waste is a difficult

  15. Numerical Zoom for Multiscale Problems with an Application to Nuclear Waste Disposal

    E-Print Network [OSTI]

    Numerical Zoom for Multiscale Problems with an Application to Nuclear Waste Disposal Jean of a nuclear waste repository site. Key words: Multiscale, Finite Element, Domain Decomposition, Chimera, Numerical Zoom, Nuclear Waste. PACS: 02.30.Jr, 47.11.Fg, 28.41.Kw, 47.55.P- 1 Introduction The present paper

  16. Introduction This paper provides the perspective of the members of the Nuclear Waste Tech-

    E-Print Network [OSTI]

    .S. program for managing spent nuclear fuel and high-level radioactive waste. It discusses the Board's opinion spent nuclear fuel. The act also established a process for evaluating the suitability of a number Waste Management Program In 1982, the U.S. Congress enacted the Nuclear Waste Policy Act (Public Law 97

  17. Scale-up considerations relevant to experimental studies of nuclear waste-package behavior

    SciTech Connect (OSTI)

    Coles, D.G.; Peters, R.D.

    1986-04-01T23:59:59.000Z

    Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

  18. Backfill composition for secondary barriers in nuclear waste repositories

    DOE Patents [OSTI]

    Beall, G.W.; Allard, B.M.

    1980-05-30T23:59:59.000Z

    A backfill composition for sorbing and retaining hazardous elements of nuclear wastes comprises 50 to 70% by weight of quartz, 10 to 30% by weight of montmorillonite, 1 to 10% by weight of phosphate mineral, 1 to 10% by weight of ferrous mineral, 1 to 10% by weight of sulfate mineral and 1 to 10% by weight of attapulgite.

  19. active nuclear wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    active nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Calculations of Induced Activity...

  20. assessing nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    assessing nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 The Initial Environmental...

  1. aqueous nuclear wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    aqueous nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Standard practice for analysis...

  2. Radioactive Waste Management in Non-Nuclear Countries - 13070

    SciTech Connect (OSTI)

    Kubelka, Dragan; Trifunovic, Dejan [SORNS, Frankopanska 11, HR-10000 Zagreb (Croatia)] [SORNS, Frankopanska 11, HR-10000 Zagreb (Croatia)

    2013-07-01T23:59:59.000Z

    This paper challenges internationally accepted concepts of dissemination of responsibilities between all stakeholders involved in national radioactive waste management infrastructure in the countries without nuclear power program. Mainly it concerns countries classified as class A and potentially B countries according to International Atomic Energy Agency. It will be shown that in such countries long term sustainability of national radioactive waste management infrastructure is very sensitive issue that can be addressed by involving regulatory body in more active way in the infrastructure. In that way countries can mitigate possible consequences on the very sensitive open market of radioactive waste management services, comprised mainly of radioactive waste generators, operators of end-life management facilities and regulatory body. (authors)

  3. Evaluation of the transport and resuspension of a simulated nuclear waste slurry: Nuclear Waste Treatment Program

    SciTech Connect (OSTI)

    Carleson, T.E.; Drown, D.C.; Hart, R.E.; Peterson, M.E.

    1987-09-01T23:59:59.000Z

    The Department of Chemical Engineering at the University of Idaho conducted research on the transport and resuspension of a simulated high-level nuclear waste slurry. In the United States, the reference process for treating both defense and civilian HLLW is vitrification using the liquid-fed ceramic melter process. The non-Newtonian behavior of the slurry complicates the evaluation of the transport and resuspension characteristics of the slurry. The resuspension of a simulated (nonradioactive) melter feed slurry was evaluated using a slurry designated as WV-205. The simulated slurry was developed for the West Valley Demonstration Project and was used during a pilot-scale ceramic melter (PSCM) experiment conducted at PNL in July 1985 (PSCM-21). This study involved determining the transport characteristics of a fully suspended slurry and the resuspension characteristics of settled solids in a pilot-scale pipe loop. The goal was to predict the transport and resuspension of a full-scale system based on rheological data for a specific slurry. The rheological behavior of the slurry was evaluated using a concentric cylinder rotational viscometer, a capillary tube viscometer, and the pilot-scale pipe loop. The results obtained from the three approaches were compared. 40 refs., 74 figs., 15 tabs.

  4. Nuclear Waste Removal Using Particle Beams Incineration with Fast Neutrons

    E-Print Network [OSTI]

    Revol, Jean Pierre Charles

    1997-01-01T23:59:59.000Z

    The management of nuclear waste is one of the major obstacles to the acceptability of nuclear power as a main source of energy for the future. TARC, a new experiment at CERN, is testing the practicality of Carlo Rubbia's idea to make use of Adiabatic Resonance Crossing to transmute long-lived fission fragments into short-lived or stable nuclides. Spallation neutrons produced in a large Lead assembly have a high probability to be captured at the energies of cross-section resonances in elements such as 99Tc, 129I, etc. An accelerator-driven sub-critical device using Thorium (Energy Amplifier) would be very effective in eliminating TRansUranic elements which constitute the most dangerous part of nuclear waste while producing from it large amounts of energy. In addition, such a system could transform, at a high rate and little energetic cost, long-lived fission fragments into short-lived elements.

  5. Technical basis for external dosimetry at the Waste Isolation Pilot Plant (WIPP)

    SciTech Connect (OSTI)

    Bradley, E.W. [Science Applications International Corp., Oak Ridge, TN (United States); Wu, C.F.; Goff, T.E. [Westinghouse Electric Corp., Carlsbad, NM (United States). Waste Isolation Div.

    1993-12-31T23:59:59.000Z

    The WIPP External Dosimetry Program, administered by Westinghouse Electric Corporation, Waste Isolation Division, for the US Department of Energy (DOE), provides external dosimetry support services for operations at the Waste Isolation Pilot Plant (WIPP) Site. These operations include the receipt, experimentation with, storage, and disposal of transuranic (TRU) wastes. This document describes the technical basis for the WIPP External Radiation Dosimetry Program. The purposes of this document are to: (1) provide assurance that the WIPP External Radiation Dosimetry Program is in compliance with all regulatory requirements, (2) provide assurance that the WIPP External Radiation Dosimetry Program is derived from a sound technical base, (3) serve as a technical reference for radiation protection personnel, and (4) aid in identifying and planning for future needs. The external radiation exposure fields are those that are documented in the WIPP Final Safety Analysis Report.

  6. 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.

    SciTech Connect (OSTI)

    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)

    2010-09-01T23:59:59.000Z

    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.

  7. Technical data summary supporting the spent nuclear fuel environment impact statement, March 1994

    SciTech Connect (OSTI)

    Geddes, R.L.; Claxton, R.E.; Lengel, J.D. [and others

    1994-03-01T23:59:59.000Z

    This report has been compiled by the WSRC Nuclear Materials Processing Division`s Planning Section at the request of the Office of Spent Fuel Management and Special Projects (EM-37) to support issuance of the Spent Nuclear Fuel Environmental Impact Statement. Savannah River Site input data evaluates five programmatic options (including {open_quotes}No Action{close_quotes}) ranging up to transfer of all DOE responsibility spent fuel to the SRS. For each option, a range of management/disposition scenarios has been examined. Each case summary provides information relative to the technical proposal, technical issues, environmental impacts, and projected costs for a forty year period (FY-35) when it is assumed that the material will be dispositioned from the SRS. The original issue of the report which was prepared under severe time constraints contained many simplifications and assumptions. Although the revisions have corrected some of the shortcomings of the original report, it is still highly recommended that significant additional study be performed before basing key decisions upon the data contained in this report. The data represents the best effort by a significant group of technical personnel familiar with nuclear materials processing, handling, and storage; but it is likely that careful scrutiny will reveal numerous discrepancies, inconsistencies and omissions. Nor does this report attempt to analyze every potential disposal pathway, but probably establishes the bounds for the most of the viable pathways. The bulk of the effort went into defining the engineering approaches necessary to execute the various mission scenarios which were changed since the last revision. The decision to limit reprocessing to only SRS aluminum clad required a major alteration of the TDS. Collection and/or calculation of much of the various waste, emission, and utility consumption data, so important to an EIS, has been updated since the last revision, but not thoroughly completed.

  8. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOE Patents [OSTI]

    Campbell, D.O.; Buxton, S.R.

    1980-06-16T23:59:59.000Z

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  9. Malonamides as new extractants for nuclear waste solutions

    SciTech Connect (OSTI)

    Cuillerdier, C.; Musikas, C.; Hoel, P.; Nigond, L.; Vitart, X. (Commissariat a L'Energie Atomique, Fontenay-aux-Roses (France))

    1991-09-01T23:59:59.000Z

    A new class of extractants has been investigated: pentaalkylpropane diamides. Due to their chelating effect on metallic cations, they extract trivalent actinides such as Am{sup 3+} and Cm{sup 3+} from acidic nitrate nuclear waste solutions. These solvents are completely incinerable and do not generate large amounts of waste. A review is provided of their chemical properties, leading to the choice of the proper molecule. The results of a bench-scale experiment performed in a mixer-settler battery are presented.

  10. Technical resource document for assured thermal processing of wastes

    SciTech Connect (OSTI)

    Farrow, R.L.; Fisk, G.A.; Hartwig, C.M.; Hurt, R.H.; Ringland, J.T.; Swansiger, W.A.

    1994-06-01T23:59:59.000Z

    This document is a concise compendium of resource material covering assured thermal processing of wastes (ATPW), an area in which Sandia aims to develop a large program. The ATPW program at Sandia is examining a wide variety of waste streams and thermal processes. Waste streams under consideration include municipal, chemical, medical, and mixed wastes. Thermal processes under consideration range from various incineration technologies to non-incineration processes such as supercritical water oxidation or molten metal technologies. Each of the chapters describes the element covered, discusses issues associated with its further development and/or utilization, presents Sandia capabilities that address these issues, and indicates important connections to other ATPW elements. The division of the field into elements was driven by the team`s desire to emphasize areas where Sandia`s capabilities can lead to major advances and is therefore somewhat unconventional. The report will be valuable to Sandians involved in further ATPW program development.

  11. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs

    Broader source: Energy.gov [DOE]

    Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs

  12. Implementing waste minimization at an active plutonium processing facility: Successes and progress at technical area (TA) -55 of the Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Balkey, J.J.; Robinson, M.A.; Boak, J.

    1997-12-01T23:59:59.000Z

    The Los Alamos National Laboratory has ongoing national security missions that necessitate increased plutonium processing. The bulk of this activity occurs at Technical Area -55 (TA-55), the nations only operable plutonium facility. TA-55 has developed and demonstrated a number of technologies that significantly minimize waste generation in plutonium processing (supercritical CO{sub 2}, Mg(OH){sub 2} precipitation, supercritical H{sub 2}O oxidation, WAND), disposition of excess fissile materials (hydride-dehydride, electrolytic decontamination), disposition of historical waste inventories (salt distillation), and Decontamination & Decommissioning (D&D) of closed nuclear facilities (electrolytic decontamination). Furthermore, TA-55 is in the process of developing additional waste minimization technologies (molten salt oxidation, nitric acid recycle, americium extraction) that will significantly reduce ongoing waste generation rates and allow volume reduction of existing waste streams. Cost savings from reduction in waste volumes to be managed and disposed far exceed development and deployment costs in every case. Waste minimization is also important because it reduces occupational exposure to ionizing radiation, risks of transportation accidents, and transfer of burdens from current nuclear operations to future generations.

  13. Technical work plan for the privatization waste characterization data quality objective process

    SciTech Connect (OSTI)

    Kirkbride, R.A.

    1996-04-24T23:59:59.000Z

    This work plan addresses the activities necessary to complete the data quality objectives process for the purpose of providing sufficient characterization information to successfully stage, pretreat, and immobilize low-activity waste per the requirements and specifications identified in the Tank Waste Remediation System Privatization Request for Proposal. The scope of this task is to complete the data quality objectives process, the results of which will provide a technical basis for sampling and characterization needs related to privatization of pretreatment and low-activity waste immobilization.

  14. Cover letter, 10/29/03, re Nuclear Safety Technical Position, Deliverable 4.2.1

    Broader source: Energy.gov [DOE]

    The enclosed Nuclear Safety Technical Position is Deliverable 4.2.1. under the Implementation Plan for Defense Nuclear Facilitises Board (DNFSB) Recommendation 2002-3, Requirements for Design...

  15. Disposal of NORM-contaminated oil field wastes in salt caverns -- Legality, technical feasibility, economics, and risk

    SciTech Connect (OSTI)

    Veil, J.A.; Smith, K.P.; Tomasko, D.; Elcock, D.; Blunt, D.; Williams, G.P.

    1998-07-01T23:59:59.000Z

    Some types of oil and gas production and processing wastes contain naturally occurring radioactive materials (NORM). If NORM is present at concentrations above regulatory levels in oil field waste, the waste requires special disposal practices. The existing disposal options for wastes containing NORM are limited and costly. This paper evaluates the legality, technical feasibility, economics, and human health risk of disposing of NORM-contaminated oil field wastes in salt caverns. Cavern disposal of NORM waste is technically feasible and poses a very low human health risk. From a legal perspective, there are no fatal flaws that would prevent a state regulatory agency from approaching cavern disposal of NORM. On the basis of the costs charged by caverns currently used for disposal of nonhazardous oil field waste (NOW), NORM waste disposal caverns could be cost competitive with existing NORM waste disposal methods when regulatory agencies approve the practice.

  16. Nuclear Waste Glasses: Beautiful Simplicity of Complex Systems

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2009-01-01T23:59:59.000Z

    The behavior of glasses with a large number of components, such as waste glasses, is not more complex than the behavior of simple glasses. On the contrary, the presence of many components restricts the composition region of these glasses in a way that allows approximating composition-property relationships by linear functions. This has far-reaching practical consequences for formulating nuclear waste glasses. On the other hand, processing high-level and low-activity waste glasses presents various problems, such as crystallization, foaming, and salt segre-gation in the melter. The need to decrease the settling of solids in the melter to an acceptable level and to maximize the rate of melting presents major challenges to processing technology. However, the most important property of the glass product is its chemical durability, a somewhat vague concept in lieu of the assessment of the glass resistance to aqueous attack while the radioactivity decays over tens of thousands of years.

  17. Radiation and Thermal Ageing of Nuclear Waste Glass

    SciTech Connect (OSTI)

    Weber, William J [ORNL

    2014-01-01T23:59:59.000Z

    The radioactive decay of fission products and actinides incorporated into nuclear waste glass leads to self-heating and self-radiation effects that may affect the stability, structure and performance of the glass in a closed system. Short-lived fission products cause significant self-heating for the first 600 years. Alpha decay of the actinides leads to self-radiation damage that can be significant after a few hundred years, and over the long time periods of geologic disposal, the accumulation of helium and radiation damage from alpha decay may lead to swelling, microstructural evolution and changes in mechanical properties. Four decades of research on the behavior of nuclear waste glass are reviewed.

  18. Alcohol-free alkoxide process for containing nuclear waste

    DOE Patents [OSTI]

    Pope, James M. (Monroeville, PA); Lahoda, Edward J. (Edgewood, PA)

    1984-01-01T23:59:59.000Z

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  19. Nuclear Waste Policy Act Signed | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /76SafeguardsSystems ModelingWaste

  20. Transmutation of nuclear waste in accelerator-driven systems

    E-Print Network [OSTI]

    Herrera-Martínez, A

    2004-01-01T23:59:59.000Z

    Today more than ever energy is not only a cornerstone of human development, but also a key to the environmental sustainability of economic activity. In this context, the role of nuclear power may be emphasized in the years to come. Nevertheless, the problems of nuclear waste, safety and proliferation still remain to be solved. It is believed that the use of accelerator-driven systems (ADSs) for nuclear waste transmutation and energy production would address these problems in a simple, clean and economically viable, and therefore sustainable, manner. This thesis covers the major nuclear physics aspects of ADSs, in particular the spallation process and the core neutronics specific to this type of systems. The need for accurate nuclear data is described, together with a detailed analysis of the specific isotopes and energy ranges in which this data needs to be improved and the impact of their uncertainty. Preliminary experimental results for some of these isotopes, produced by the Neutron Time-of-Flight (n_TOF) ...

  1. Nuclear Waste Policy Act.doc

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates fromNuclear Security

  2. Nuclear waste vitrification efficiency: Cold cap reactions

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecoveryNuclearLife

  3. Consideration of nuclear criticality when disposing of transuranic waste at the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    RECHARD,ROBERT P.; SANCHEZ,LAWRENCE C.; STOCKMAN,CHRISTINE T.; TRELLUE,HOLLY R.

    2000-04-01T23:59:59.000Z

    Based on general arguments presented in this report, nuclear criticality was eliminated from performance assessment calculations for the Waste Isolation Pilot Plant (WIPP), a repository for waste contaminated with transuranic (TRU) radioisotopes, located in southeastern New Mexico. At the WIPP, the probability of criticality within the repository is low because mechanisms to concentrate the fissile radioisotopes dispersed throughout the waste are absent. In addition, following an inadvertent human intrusion into the repository (an event that must be considered because of safety regulations), the probability of nuclear criticality away from the repository is low because (1) the amount of fissile mass transported over 10,000 yr is predicted to be small, (2) often there are insufficient spaces in the advective pore space (e.g., macroscopic fractures) to provide sufficient thickness for precipitation of fissile material, and (3) there is no credible mechanism to counteract the natural tendency of the material to disperse during transport and instead concentrate fissile material in a small enough volume for it to form a critical concentration. Furthermore, before a criticality would have the potential to affect human health after closure of the repository--assuming that a criticality could occur--it would have to either (1) degrade the ability of the disposal system to contain nuclear waste or (2) produce significantly more radioisotopes than originally present. Neither of these situations can occur at the WIPP; thus, the consequences of a criticality are also low.

  4. Method of determining a content of a nuclear waste container

    DOE Patents [OSTI]

    Bernardi, Richard T. (Prospect Heights, IL); Entwistle, David (Buffalo Grove, IL)

    2003-04-22T23:59:59.000Z

    A method and apparatus are provided for identifying contents of a nuclear waste container. The method includes the steps of forming an image of the contents of the container using digital radiography, visually comparing contents of the image with expected contents of the container and performing computer tomography on the container when the visual inspection reveals an inconsistency between the contents of the image and the expected contents of the container.

  5. Potential applications of nanostructured materials in nuclear waste management.

    SciTech Connect (OSTI)

    Braterman, Paul S. (The University of North Texas, Denton, TX); Phol, Phillip Isabio; Xu, Zhi-Ping (The University of North Texas, Denton, TX); Brinker, C. Jeffrey; Yang, Yi (University of New Mexico, Albuquerque, NM); Bryan, Charles R.; Yu, Kui; Xu, Huifang (University of New Mexico, Albuquerque, NM); Wang, Yifeng; Gao, Huizhen

    2003-09-01T23:59:59.000Z

    This report summarizes the results obtained from a Laboratory Directed Research & Development (LDRD) project entitled 'Investigation of Potential Applications of Self-Assembled Nanostructured Materials in Nuclear Waste Management'. The objectives of this project are to (1) provide a mechanistic understanding of the control of nanometer-scale structures on the ion sorption capability of materials and (2) develop appropriate engineering approaches to improving material properties based on such an understanding.

  6. Information basis for developing comprehensive waste management system-US-Japan joint nuclear energy action plan waste management working group phase I report.

    SciTech Connect (OSTI)

    Nutt, M.; Nuclear Engineering Division

    2010-05-25T23:59:59.000Z

    The activity of Phase I of the Waste Management Working Group under the United States - Japan Joint Nuclear Energy Action Plan started in 2007. The US-Japan JNEAP is a bilateral collaborative framework to support the global implementation of safe, secure, and sustainable, nuclear fuel cycles (referred to in this document as fuel cycles). The Waste Management Working Group was established by strong interest of both parties, which arise from the recognition that development and optimization of waste management and disposal system(s) are central issues of the present and future nuclear fuel cycles. This report summarizes the activity of the Waste Management Working Group that focused on consolidation of the existing technical basis between the U.S. and Japan and the joint development of a plan for future collaborative activities. Firstly, the political/regulatory frameworks related to nuclear fuel cycles in both countries were reviewed. The various advanced fuel cycle scenarios that have been considered in both countries were then surveyed and summarized. The working group established the working reference scenario for the future cooperative activity that corresponds to a fuel cycle scenario being considered both in Japan and the U.S. This working scenario involves transitioning from a once-through fuel cycle utilizing light water reactors to a one-pass uranium-plutonium fuel recycle in light water reactors to a combination of light water reactors and fast reactors with plutonium, uranium, and minor actinide recycle, ultimately concluding with multiple recycle passes primarily using fast reactors. Considering the scenario, current and future expected waste streams, treatment and inventory were discussed, and the relevant information was summarized. Second, the waste management/disposal system optimization was discussed. Repository system concepts were reviewed, repository design concepts for the various classifications of nuclear waste were summarized, and the factors to consider in repository design and optimization were then discussed. Japan is considering various alternatives and options for the geologic disposal facility and the framework for future analysis of repository concepts was discussed. Regarding the advanced waste and storage form development, waste form technologies developed in both countries were surveyed and compared. Potential collaboration areas and activities were next identified. Disposal system optimization processes and techniques were reviewed, and factors to consider in future repository design optimization activities were also discussed. Then the potential collaboration areas and activities related to the optimization problem were extracted.

  7. Preliminary technical and legal evaluation of disposing of nonhazardous oil field waste into salt caverns

    SciTech Connect (OSTI)

    Veil, J.; Elcock, D.; Raivel, M.; Caudle, D.; Ayers, R.C. Jr.; Grunewald, B.

    1996-06-01T23:59:59.000Z

    Caverns can be readily formed in salt formations through solution mining. The caverns may be formed incidentally, as a result of salt recovery, or intentionally to create an underground chamber that can be used for storing hydrocarbon products or compressed air or disposing of wastes. The purpose of this report is to evaluate the feasibility, suitability, and legality of disposing of nonhazardous oil and gas exploration, development, and production wastes (hereafter referred to as oil field wastes, unless otherwise noted) in salt caverns. Chapter 2 provides background information on: types and locations of US subsurface salt deposits; basic solution mining techniques used to create caverns; and ways in which salt caverns are used. Later chapters provide discussion of: federal and state regulatory requirements concerning disposal of oil field waste, including which wastes are considered eligible for cavern disposal; waste streams that are considered to be oil field waste; and an evaluation of technical issues concerning the suitability of using salt caverns for disposing of oil field waste. Separate chapters present: types of oil field wastes suitable for cavern disposal; cavern design and location; disposal operations; and closure and remediation. This report does not suggest specific numerical limits for such factors or variables as distance to neighboring activities, depths for casings, pressure testing, or size and shape of cavern. The intent is to raise issues and general approaches that will contribute to the growing body of information on this subject.

  8. Yucca Mountain: How Do Global and Federal Initiatives Impact Clark County's Nuclear Waste Program?

    SciTech Connect (OSTI)

    Navis, I.; McGehee, B. [Clark County Department of Comprehensive Planning - Nuclear Waste Division, Las Vegas, NV (United States)

    2008-07-01T23:59:59.000Z

    Since 1987, Clark County has been designated by the U.S. Department of Energy (DOE) as an 'Affected Unit of Local Government' (AULG). The AULG designation is an acknowledgement by the federal government that activities associated with the Yucca Mountain proposal could result in considerable impacts on Clark County residents and the community as a whole. As an AULG, Clark County is authorized to identify 'any potential economic, social, public health and safety, and environmental impacts of a repository', 42 U.S.C. Section 10135(c)(1)(B)(i) under provisions of the Nuclear Waste Policy Act Amendments (NWPAA). Clark County's oversight program contains key elements of (1) technical and scientific analysis (2) transportation analysis (3) impact assessment and monitoring (4) policy and legislative analysis and monitoring, and (5) public outreach. Clark County has conducted numerous studies of potential impacts, many of which are summarized in Clark County's Impact Assessment Report that was submitted DOE and the President of the United States in February 2002. Given the unprecedented magnitude and duration of DOE's proposal, as well as the many unanswered questions about the transportation routes, number of shipments, and the modal mix that will ultimately be used, impacts to public health and safety and security, as well as socioeconomic impacts, can only be estimated. In order to refine these estimates, Clark County Comprehensive Planning Department's Nuclear Waste Division updates, assesses, and monitors impacts on a regular basis. Clark County's Impact Assessment program covers not only unincorporated Clark County but all five jurisdictions of Las Vegas, North Las Vegas, Henderson, Mesquite, and Boulder City as well as tribal jurisdictions that fall within Clark County's geographic boundary. National and global focus on nuclear power and nuclear waste could have significant impact on the Yucca Mountain Program, and therefore, Clark County's oversight of that program. (authors)

  9. Analysis of waste treatment requirements for DOE mixed wastes: Technical basis

    SciTech Connect (OSTI)

    NONE

    1995-02-01T23:59:59.000Z

    The risks and costs of managing DOE wastes are a direct function of the total quantities of 3wastes that are handled at each step of the management process. As part of the analysis of the management of DOE low-level mixed wastes (LLMW), a reference scheme has been developed for the treatment of these wastes to meet EPA criteria. The treatment analysis in a limited form was also applied to one option for treatment of transuranic wastes. The treatment requirements in all cases analyzed are based on a reference flowsheet which provides high level treatment trains for all LLMW. This report explains the background and basis for that treatment scheme. Reference waste stream chemical compositions and physical properties including densities were established for each stream in the data base. These compositions are used to define the expected behavior for wastes as they pass through the treatment train. Each EPA RCRA waste code was reviewed, the properties, chemical composition, or characteristics which are of importance to waste behavior in treatment were designated. Properties that dictate treatment requirements were then used to develop the treatment trains and identify the unit operations that would be included in these trains. A table was prepared showing a correlation of the waste physical matrix and the waste treatment requirements as a guide to the treatment analysis. The analysis of waste treatment loads is done by assigning wastes to treatment steps which would achieve RCRA compliant treatment. These correlation`s allow one to examine the treatment requirements in a condensed manner and to see that all wastes and contaminant sets are fully considered.

  10. Technical and Political Assessment of Peaceful Nuclear Power Program Prospects in North Africa and the

    E-Print Network [OSTI]

    PNNL-16840 Technical and Political Assessment of Peaceful Nuclear Power Program Prospects in North and application of resources towards developing nuclear-generated electricity and nuclear-powered desalination nations have recently expressed interest in developing nuclear energy for peaceful purposes. Many

  11. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    SciTech Connect (OSTI)

    Schultz, Peter Andrew

    2011-12-01T23:59:59.000Z

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  12. Removal of Pertechnetate from Simulated Nuclear Waste Streams Using Supported Zerovalent Iron

    E-Print Network [OSTI]

    Removal of Pertechnetate from Simulated Nuclear Waste Streams Using Supported Zerovalent Iron John and reduction of pertechnetate anions (TcO4 -) from complex waste mixtures was investigated as an alternative approach to current waste-processing schemes. Although applicable to pertechnetate-containing waste streams

  13. Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste

    SciTech Connect (OSTI)

    Turick, C; Berry, C.

    2012-10-15T23:59:59.000Z

    Long-term storage of low level radioactive material in below ground concrete disposal units (DUs) (Saltstone Disposal Facility) is a means of depositing wastes generated from nuclear operations of the U.S. Department of Energy. Based on the currently modeled degradation mechanisms, possible microbial induced effects on the structural integrity of buried low level wastes must be addressed. Previous international efforts related to microbial impacts on concrete structures that house low level radioactive waste showed that microbial activity can play a significant role in the process of concrete degradation and ultimately structural deterioration. This literature review examines the recent research in this field and is focused on specific parameters that are applicable to modeling and prediction of the fate of concrete vaults housing stored wastes and the wastes themselves. Rates of concrete biodegradation vary with the environmental conditions, illustrating a need to understand the bioavailability of key compounds involved in microbial activity. Specific parameters require pH and osmotic pressure to be within a certain range to allow for microbial growth as well as the availability and abundance of energy sources like components involved in sulfur, iron and nitrogen oxidation. Carbon flow and availability are also factors to consider in predicting concrete biodegradation. The results of this review suggest that microbial activity in Saltstone, (grouted low level radioactive waste) is unlikely due to very high pH and osmotic pressure. Biodegradation of the concrete vaults housing the radioactive waste however, is a possibility. The rate and degree of concrete biodegradation is dependent on numerous physical, chemical and biological parameters. Results from this review point to parameters to focus on for modeling activities and also, possible options for mitigation that would minimize concrete biodegradation. In addition, key chemical components that drive microbial activity on concrete surfaces are discussed.

  14. Nuclear waste repository transparency technology test bed demonstrations at WIPP

    SciTech Connect (OSTI)

    BETSILL,J. DAVID; ELKINS,NED Z.; WU,CHUAN-FU; MEWHINNEY,JAMES D.; AAMODT,PAUL

    2000-01-27T23:59:59.000Z

    Secretary of Energy, Bill Richardson, has stated that one of the nuclear waste legacy issues is ``The challenge of managing the fuel cycle's back end and assuring the safe use of nuclear power.'' Waste management (i.e., the back end) is a domestic and international issue that must be addressed. A key tool in gaining acceptance of nuclear waste repository technologies is transparency. Transparency provides information to outside parties for independent assessment of safety, security, and legitimate use of materials. Transparency is a combination of technologies and processes that apply to all elements of the development, operation, and closure of a repository system. A test bed for nuclear repository transparency technologies has been proposed to develop a broad-based set of concepts and strategies for transparency monitoring of nuclear materials at the back end of the fuel/weapons cycle. WIPP is the world's first complete geologic repository system for nuclear materials at the back end of the cycle. While it is understood that WIPP does not currently require this type of transparency, this repository has been proposed as realistic demonstration site to generate and test ideas, methods, and technologies about what transparency may entail at the back end of the nuclear materials cycle, and which could be applicable to other international repository developments. An integrated set of transparency demonstrations was developed and deployed during the summer, and fall of 1999 as a proof-of-concept of the repository transparency technology concept. These demonstrations also provided valuable experience and insight into the implementation of future transparency technology development and application. These demonstrations included: Container Monitoring Rocky Flats to WIPP; Underground Container Monitoring; Real-Time Radiation and Environmental Monitoring; Integrated level of confidence in the system and information provided. As the world's only operating deep geologic repository, the Waste Isolation Pilot Plant (WIPP) offers a unique opportunity to serve as an international cooperative test bed for developing and demonstrating technologies and processes in a fully operational repository system setting. To address the substantial national security implications for the US resulting from the lack of integrated, transparent management and disposition of nuclear materials at the back-end of the nuclear fuel and weapons cycles, it is proposed that WIPP be used as a test bed to develop and demonstrate technologies that will enable the transparent and proliferation-resistant geologic isolation of nuclear materials. The objectives of this initiative are to: (1) enhance public confidence in safe, secure geologic isolation of nuclear materials; (2) develop, test, and demonstrate transparency measures and technologies for the back-end of nuclear fuel cycle; and (3) foster international collaborations leading to workable, effective, globally-accepted standards for the transparent monitoring of geological repositories for nuclear materials. Test-bed activities include: development and testing of monitoring measures and technologies; international demonstration experiments; transparency workshops; visiting scientist exchanges; and educational outreach. These activities are proposed to be managed by the Department of Energy/Carlsbad Area Office (DOE/CAO) as part of The Center for Applied Repository and Underground Studies (CARUS).

  15. Technical baseline description of high-level waste andlow-activity waste feed mobilization and delivery

    SciTech Connect (OSTI)

    Papp, I.G. [Numatec Hanford Corporation, Richland, WA 99352 (United States)

    1997-06-01T23:59:59.000Z

    This document is a compilation of information related to the high-level waste (HLW) and low-activity waste (LAW) feed staging, mobilization, and transfer/delivery issues. Information relevant to current Tank Waste Remediation System (TWRS) inventories and activities designed to feed the Phase I Privatization effort at the Hanford Site is included. Discussions on the higher level Phase II activities are offered for a perspective on the interfaces.

  16. Conceivable new recycling of nuclear waste by nuclear power companies in their plants

    E-Print Network [OSTI]

    Ruggero Maria Santilli

    1997-04-09T23:59:59.000Z

    We outline the basic principles and the needed experiments for a conceivable new recycling of nuclear waste by the power plants themselves to avoid its transportation and storage to a (yet unknown) dumping area. Details are provided in an adjoining paper and in patents pending.

  17. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository - BMT1 of the DECOVALEX III project. Part 3: Effects of THM coupling in sparsely fractured rocks

    E-Print Network [OSTI]

    2004-01-01T23:59:59.000Z

    Safety of a Hypothetical Nuclear Waste Repository – BMT1 ofsafety of a hypothetical nuclear waste repository – BMT1 ofsafety of a hypothetical nuclear waste repository – BMT1 of

  18. Technical area status report for low-level mixed waste final waste forms. Volume 2, Appendices

    SciTech Connect (OSTI)

    Mayberry, J.L.; Huebner, T.L. [Science Applications International Corp., Idaho Falls, ID (United States); Ross, W. [Pacific Northwest Lab., Richland, WA (United States); Nakaoka, R. [Los Alamos National Lab., NM (United States); Schumacher, R. [Westinghouse Savannah River Co., Aiken, SC (United States); Cunnane, J.; Singh, D. [Argonne National Lab., IL (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Greenhalgh, W. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-08-01T23:59:59.000Z

    This report presents information on low-level mixed waste forms.The descriptions of the low-level mixed waste (LLMW) streams that are considered by the Mixed Waste Integrated Program (MWIP) are given in Appendix A. This information was taken from descriptions generated by the Mixed Waste Treatment Program (MWTP). Appendix B provides a list of characteristic properties initially considered by the Final Waste Form (FWF) Working Group (WG). A description of facilities available to test the various FWFs discussed in Volume I of DOE/MWIP-3 are given in Appendix C. Appendix D provides a summary of numerous articles that were reviewed on testing of FWFS. Information that was collected by the tests on the characteristic properties considered in this report are documented in Appendix D. The articles reviewed are not a comprehensive list, but are provided to give an indication of the data that are available.

  19. Sandia Energy - Waste Isolation Pilot Plant Technical Assessment Report

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's PossibleRadiationImplementing Nonlinear757KelleyEffectsonSandia'sEvent VideoWasteReleased

  20. Savannah River Site - Salt Waste Processing Facility Independent Technical Review

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmentalEfficiency | Department ofEnergySaraKSALT WASTE

  1. NWTS program criteria for mined geologic disposal of nuclear waste: program objectives, functional requirements, and system performance criteria

    SciTech Connect (OSTI)

    None

    1981-04-01T23:59:59.000Z

    At the present time, final repository criteria have not been issued by the responsible agencies. This document describes general objectives, requirements, and criteria that the DOE intends to apply in the interim to the National Waste Terminal Storage (NWTS) Program. These objectives, requirements, and criteria have been developed on the basis of DOE's analysis of what is needed to achieve the National objective of safe waste disposal in an environmentally acceptable and economic manner and are expected to be consistent with anticipated regulatory standards. The qualitative statements in this document address the broad issues of public and occupational health and safety, institutional acceptability, engineering feasibility, and economic considerations. A comprehensive set of criteria, general and project specific, of which these are a part, will constitute a portion of the technical basis for preparation and submittal by the DOE of formal documents to support future license applications for nuclear waste repositories.

  2. Advanced Mixed Waste Treatment Project melter system preliminary design technical review meeting

    SciTech Connect (OSTI)

    Eddy, T.L.; Raivo, B.D.; Soelberg, N.R.; Wiersholm, O.

    1995-02-01T23:59:59.000Z

    The Idaho National Engineering Laboratory Advanced Mixed Waste Treatment Project sponsored a plasma are melter technical design review meeting to evaluate high-temperature melter system configurations for processing heterogeneous alpha-contaminated low-level radioactive waste (ALLW). Thermal processing experts representing Department of Energy contractors, the Environmental Protection Agency, and private sector companies participated in the review. The participants discussed issues and evaluated alternative configurations for three areas of the melter system design: plasma torch melters and graphite arc melters, offgas treatment options, and overall system configuration considerations. The Technical Advisory Committee for the review concluded that graphite arc melters are preferred over plasma torch melters for processing ALLW. Initiating involvement of stakeholders was considered essential at this stage of the design. For the offgas treatment system, the advisory committee raised the question whether to a use wet-dry or a dry-wet system. The committee recommended that the waste stream characterization, feed preparation, and the control system are essential design tasks for the high-temperature melter treatment system. The participants strongly recommended that a complete melter treatment system be assembled to conduct tests with nonradioactive surrogate waste material. A nonradioactive test bed would allow for inexpensive design and operational changes prior to assembling a system for radioactive waste treatment operations.

  3. Selected, annotated bibliography of studies relevant to the isolation of nuclear wastes. [705 references

    SciTech Connect (OSTI)

    Hyder, L.K.; Fore, C.S.; Vaughan, N.D.; Faust, R.A.

    1980-09-01T23:59:59.000Z

    This annotated bibliography of 705 references represents the first in a series to be published by the Ecological Sciences Information Center containing scientific, technical, economic, and regulatory information relevant to nuclear waste isolation. Most references discuss deep geologic disposal, with fewer studies of deep seabed disposal; space disposal is also included. The publication covers both domestic and foreign literature for the period 1954 to 1980. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Envirnmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Repository Design and Engineering; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. When the author is not given, the corporate affiliation appears first. If these two levels of authorship are not given, the title of the document is used as the identifying level. Indexes are provided for author(s), keywords, subject category, title, geographic location, measured parameters, measured radionuclides, and publication description.

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

    SciTech Connect (OSTI)

    Vienna, John D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Ryan, Joseph V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Gin, Stephane [CEA Marcoule, DTCD SECM, Bagnols-sur-Ceze (France); Inagaki, Yaohiro [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Fukuoda (Japan)

    2013-12-01T23:59:59.000Z

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

  5. Cesium and Strontium Specific Exchangers for Nuclear Waste Effluent Remediation

    SciTech Connect (OSTI)

    A. Clearfield; A. I. Bortun; L. A. Bortun; E. A. Bhlume; P. Sylvester; G. M. Graziano

    2000-09-01T23:59:59.000Z

    During the past 50 years, nuclear defense activities have produced large quantities of nuclear waste that now require safe and permanent disposal. The general procedure to be implemented involves the removal of cesium and strontium from the waste solutions for disposal in permanently vitrified media. This requires highly selective sorbents or ion exchangers. Further, at the high radiation doses present in the solution, organic exchangers or sequestrants are likely to decompose over time. Inorganic ion exchangers are resistant to radiation damage and can exhibit remarkably high selectivities. We have synthesized three families of tunnel-type ion exchangers. The crystal structures of these compounds as well as their protonated phases, coupled with ion exchange titrations, were determined and this information was used to develop an understanding of their ion exchange behavior. The ion exchange selectivities of these phases could be regulated by isomorphous replacement of the framework metals by larger or smaller radius metals. In the realm of layered compounds, we prepared alumina, silica, and zirconia pillared clays and sodium micas. The pillared clays yielded very high Kd values for Cs+ and were very effective in removing Cs+ from groundwaters. The sodium micas also had a high affinity for Cs+ but an even greater attraction for S42+. They also possess the property of trapping these ions permanently as the layers slowly decrease their interlayer distance as loading occurs. Sodium nonatitanate exhibited extremely high Kd values for Sr2+ in alkaline tank wastes and should be considered for removal of Sr2+ in such cases. For tank wastes containing complexing agents, we have found that adding Ca2+ to the solution releases the complexed Sr2+ which may then be removed with the CST exchanger.

  6. Flammability Control In A Nuclear Waste Vitrification System

    SciTech Connect (OSTI)

    Zamecnik, John R.; Choi, Alexander S.; Johnson, Fabienne C.; Miller, Donald H.; Lambert, Daniel P.; Stone, Michael E.; Daniel, William E. Jr.

    2013-07-25T23:59:59.000Z

    The Defense Waste Processing Facility at the Savannah River Site processes high-level radioactive waste from the processing of nuclear materials that contains dissolved and precipitated metals and radionuclides. Vitrification of this waste into borosilicate glass for ultimate disposal at a geologic repository involves chemically modifying the waste to make it compatible with the glass melter system. Pretreatment steps include removal of excess aluminum by dissolution and washing, and processing with formic and nitric acids to: 1) adjust the reduction-oxidation (redox) potential in the glass melter to reduce radionuclide volatility and improve melt rate; 2) adjust feed rheology; and 3) reduce by steam stripping the amount of mercury that must be processed in the melter. Elimination of formic acid in pretreatment has been studied to eliminate the production of hydrogen in the pretreatment systems, which requires nuclear grade monitoring equipment. An alternative reductant, glycolic acid, has been studied as a substitute for formic acid. However, in the melter, the potential for greater formation of flammable gases exists with glycolic acid. Melter flammability is difficult to control because flammable mixtures can be formed during surges in offgases that both increase the amount of flammable species and decrease the temperature in the vapor space of the melter. A flammable surge can exceed the 60% of the LFL with no way to mitigate it. Therefore, careful control of the melter feed composition based on scaled melter surge testing is required. The results of engineering scale melter tests with the formic-nitric flowsheet and the use of these data in the melter flammability model are presented.

  7. CHARACTERIZATION OF DEFENSE NUCLEAR WASTE USING HAZARDOUS WASTE GUIDANCE. APPLICATIONS TO HANFORD SITE ACCELERATED HIGH-LEVEL WASTE TREATMENT AND DISPOSAL MISSION0

    SciTech Connect (OSTI)

    Hamel, William; Huffman, Lori; Lerchen, Megan; Wiemers, Karyn

    2003-02-27T23:59:59.000Z

    Federal hazardous waste regulations were developed for management of industrial waste. These same regulations are also applicable for much of the nation's defense nuclear wastes. At the U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State, one of the nation's largest inventories of nuclear waste remains in storage in large underground tanks. The waste's regulatory designation and its composition and form constrain acceptable treatment and disposal options. Obtaining detailed knowledge of the tank waste composition presents a significant portion of the many challenges in meeting the regulatory-driven treatment and disposal requirements for this waste. Key in applying the hazardous waste regulations to defense nuclear wastes is defining the appropriate and achievable quality for waste feed characterization data and the supporting evidence demonstrating that applicable requirements have been met at the time of disposal. Application of a performance-based approach to demonstrating achievable quality standards will be discussed in the context of the accelerated high-level waste treatment and disposal mission at the Hanford Site.

  8. THERMODYNAMIC TABLES FOR NUCLEAR WASTE ISOLATION, V.1: AQUEOUSSOLUTIONS DATABASE

    SciTech Connect (OSTI)

    Phillips, S.L.; Hale, F.V.; Silvester, L.F.

    1988-05-01T23:59:59.000Z

    Tables of consistent thermodynamic property values for nuclear waste isolation are given. The tables include critically assessed values for Gibbs energy of formation. enthalpy of formation, entropy and heat capacity for minerals; solids; aqueous ions; ion pairs and complex ions of selected actinide and fission decay products at 25{sup o}C and zero ionic strength. These intrinsic data are used to calculate equilibrium constants and standard potentials which are compared with typical experimental measurements and other work. Recommendations for additional research are given.

  9. Spectroscopic investigation of simulated low-level nuclear waste glass

    SciTech Connect (OSTI)

    Rong, Chaoying; Li, Hong; Hrma, P.R.; Cho, H.M. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-12-31T23:59:59.000Z

    Borosilicate glasses with high sodium concentrations, formulated to simulate vitrified Hanford low-level wastes (LLW), were investigated by {sup 31}P magic angle spinning (MAS) nuclear magnetic resonance (NMR). Phase separation, glass homogeneity changes during remelting, and the form of phosphate in glass following product consistency tests (PCT) were also examined by NMR. The results show that a distinct orthophosphate phase not part of the glass network is present in the glass. The effect of glass composition on phosphate chemical environments in the glass is discussed.

  10. Nuclear Waste Fund Activities Management Team | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates fromNuclear Security ConferenceWaste

  11. Low level mixed waste thermal treatment technical basis report

    SciTech Connect (OSTI)

    Place, B.G.

    1994-12-01T23:59:59.000Z

    Detailed characterization of the existing and projected Hanford Site Radioactive Mixed Waste (RMW) inventory was initiated in 1993 (Place 1993). This report presents an analysis of the existing and projected RMW inventory. The subject characterization effort continues to be in support of the following engineering activities related to thermal treatment of Hanford Site RMW: (1) Contracting for commercial thermal treatment; (2) Installation and operation of an onsite thermal treatment facility (Project W-242); (3) Treatment at another Department of Energy (DOE) site. The collation of this characterization information (data) has emphasized the establishment of a common data base for the entire existing RMW inventory so that the specification of feed streams destined for different treatment facilities can be coordinated.

  12. Nuclear waste management. Semiannual progress report, April 1983-September 1983

    SciTech Connect (OSTI)

    McElroy, J.L.; Powell, J.A. (comps.)

    1984-01-01T23:59:59.000Z

    The status of the following programs is reported: waste stabilization; waste isolation; low-level waste management; remedial action; and supporting studies. 58 figures, 39 tables.

  13. Characterising encapsulated nuclear waste using cosmic-ray muon tomography

    E-Print Network [OSTI]

    Clarkson, Anthony; Hoek, Matthias; Ireland, David G; Johnstone, John R; Kaiser, Ralf; Keri, Tibor; Lumsden, Scott; Mahon, David F; McKinnon, Bryan; Murray, Morgan; Nutbeam-Tuffs, Siân; Shearer, Craig; Yang, Guangliang; Zimmerman, Colin

    2014-01-01T23:59:59.000Z

    Tomographic imaging techniques using the Coulomb scattering of cosmic-ray muons have been shown previously to successfully identify and characterise low- and high-Z materials within an air matrix using a prototype scintillating-fibre tracker system. Those studies were performed as the first in a series to assess the feasibility of this technology and image reconstruction techniques in characterising the potential high-Z contents of legacy nuclear waste containers for the UK Nuclear Industry. The present work continues the feasibility study and presents the first images reconstructed from experimental data collected using this small-scale prototype system of low- and high-Z materials encapsulated within a concrete-filled stainless-steel container. Clear discrimination is observed between the thick steel casing, the concrete matrix and the sample materials assayed. These reconstructed objects are presented and discussed in detail alongside the implications for future industrial scenarios.

  14. Characterising encapsulated nuclear waste using cosmic-ray muon tomography

    E-Print Network [OSTI]

    Anthony Clarkson; David J. Hamilton; Matthias Hoek; David G. Ireland; John R. Johnstone; Ralf Kaiser; Tibor Keri; Scott Lumsden; David F. Mahon; Bryan McKinnon; Morgan Murray; Siân Nutbeam-Tuffs; Craig Shearer; Guangliang Yang; Colin Zimmerman

    2014-10-27T23:59:59.000Z

    Tomographic imaging techniques using the Coulomb scattering of cosmic-ray muons have been shown previously to successfully identify and characterise low- and high-Z materials within an air matrix using a prototype scintillating-fibre tracker system. Those studies were performed as the first in a series to assess the feasibility of this technology and image reconstruction techniques in characterising the potential high-Z contents of legacy nuclear waste containers for the UK Nuclear Industry. The present work continues the feasibility study and presents the first images reconstructed from experimental data collected using this small-scale prototype system of low- and high-Z materials encapsulated within a concrete-filled stainless-steel container. Clear discrimination is observed between the thick steel casing, the concrete matrix and the sample materials assayed. These reconstructed objects are presented and discussed in detail alongside the implications for future industrial scenarios.

  15. Nuclear Waste Technical Review Board Members: Curricula Vitae

    E-Print Network [OSTI]

    , and private organizations, including the White House, Department of Energy, National Academy of Sciences and societies. In particular, he has served as president of the Ecological Society of America; president ecology, micro-environments, Alaskan tundra vegetation, and academic administration and research related

  16. U.S. Nuclear Waste Technical Review Board

    E-Print Network [OSTI]

    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .33 Sweden and Finland

  17. Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked QuestionsDepartment ofDepartment640 FederalDepartment83-2007

  18. Branch technical position on the use of expert elicitation in the high-level radioactive waste program

    SciTech Connect (OSTI)

    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)

    1996-11-01T23:59:59.000Z

    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.

  19. Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

    SciTech Connect (OSTI)

    Not Available

    1980-10-01T23:59:59.000Z

    Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

  20. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    SciTech Connect (OSTI)

    Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

    2014-10-02T23:59:59.000Z

    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 multiphase 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.

  1. Role of inorganic chemistry on nuclear energy examined

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    examined Inorganic chemistry can provide insight and improve technical issues surrounding nuclear power production and waste disposition. July 31, 2013 Aspects of inorganic...

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

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19T23:59:59.000Z

    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.

  3. Separation of technetium from nuclear waste stream simulants. Final report

    SciTech Connect (OSTI)

    Strauss, S.H. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Chemistry

    1995-09-11T23:59:59.000Z

    The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering {sup 99}TcO{sub 4}{sup {minus}} from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO{sub 4}{sup {minus}}), a stable (non-radioactive) chemical surrogate for {sup 99}TcO{sub 4}{sup {minus}}. Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO{sub 4}{sup {minus}} and TcO{sub 4}{sup {minus}}.

  4. Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

    E-Print Network [OSTI]

    Mueller, Alex C

    2012-01-01T23:59:59.000Z

    While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven syste...

  5. Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

    E-Print Network [OSTI]

    Alex C. Mueller

    2012-10-16T23:59:59.000Z

    While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

  6. Disposition of nuclear waste using subcritical accelerator-driven systems

    SciTech Connect (OSTI)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-12-31T23:59:59.000Z

    Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power.

  7. Corrosion of Nuclear Fuel Inside a Failed Copper Nuclear Waste Container

    SciTech Connect (OSTI)

    Broczkowski, Michael E.; Goldik, Jonathan S.; Santos, Billy G.; Noel, James J. [University of Western Ontario, London, ON, N6A 5B7 (Canada); Shoesmith, David [Chemistry, University of Western Ontario, Chemistry Building, London, ON, N6A 5B7 (Canada)

    2007-07-01T23:59:59.000Z

    Canada's Nuclear Waste Management Organization has recommended to the Canadian federal government an adaptive phased management approach to the long-term management of used nuclear fuel. This approach includes isolation in a deep geologic repository. In such a repository, the fuel would be sealed inside a carbon steel-lined copper container. To assist the development of performance assessment models studies of fuel behaviour inside a failed waste container are underway. Using an iterative modeling and experimental approach, the important features and processes that determine fuel behaviour have been identified and studied. These features and processes are discussed and the results of studies to elucidate specific mechanisms and determine important parameter values summarized. (authors)

  8. Precipitation-adsorption process for the decontamination of nuclear waste supernates

    DOE Patents [OSTI]

    Lee, Lien-Mow (North Augusta, SC); Kilpatrick, Lester L. (Aiken, SC)

    1984-01-01T23:59:59.000Z

    High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

  9. Precipitation-adsorption process for the decontamination of nuclear waste supernates

    DOE Patents [OSTI]

    Lee, L.M.; Kilpatrick, L.L.

    1982-05-19T23:59:59.000Z

    High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

  10. Federal Technical Capability Policy for Defense Nuclear Facilities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1998-12-10T23:59:59.000Z

    The policy regarding the Federal Technical Capability Program, which provides for the recruitment, deployment, development, and retention of Federal personnel with the demonstrated technical capability to safely accomplish the Department’s missions and responsibilities.

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

    SciTech Connect (OSTI)

    M. D. Staiger

    2007-06-01T23:59:59.000Z

    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.

  12. Standard Guide for Preparing Waste Management Plans for Decommissioning Nuclear Facilities

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This guide addresses the development of waste management plans for potential waste streams resulting from decommissioning activities at nuclear facilities, including identifying, categorizing, and handling the waste from generation to final disposal. 1.2 This guide is applicable to potential waste streams anticipated from decommissioning activities of nuclear facilities whose operations were governed by the Nuclear Regulatory Commission (NRC) or Agreement State license, under Department of Energy (DOE) Orders, or Department of Defense (DoD) regulations. 1.3 This guide provides a description of the key elements of waste management plans that if followed will successfully allow for the characterization, packaging, transportation, and off-site treatment or disposal, or both, of conventional, hazardous, and radioactive waste streams. 1.4 This guide does not address the on-site treatment, long term storage, or on-site disposal of these potential waste streams. 1.5 This standard does not purport to address ...

  13. Technical Basis for the Determination that Current Characterization Data and Processes are Sufficient to Ensure Safe Storage and to Design Waste Disposal

    SciTech Connect (OSTI)

    SIMPSON, B.C.

    1999-08-12T23:59:59.000Z

    This document presents the technical basis for closure of Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-5 Implementation Plan milestone 5.6.3.13, ''Core sample all tanks by 2002'' (DOE-RL 1996). The milestone was based on the need for characterization data to ensure safe storage of the waste, to operate the tanks safely, and to plan and implement retrieval and processing of the waste. Sufficient tank characterization data have been obtained to ensure that existing controls are adequate for safe storage of the waste in the 177 waste tanks at the Hanford Site. In addition, a process has been developed, executed, and institutionalized to systemically identify information needs, to integrate and prioritize the needs, and to reliably obtain and analyze the associated samples. This document provides a technical case that the remaining 45 incompletely sampled tanks no longer require sampling to support the intent of the Implementation Plan milestone. Sufficient data have been obtained to close the Unreviewed Safety Questions (USQs), and to ensure that existing hazard controls are adequate and appropriately applied. However, in the future, additional characterization of tanks at the site will be required to support identified information needs. Closure of this milestone allows sampling and analytical data to be obtained in a manner that is consistent with the integrated priority process.

  14. Senior Technical Safety Manager Qualification Program Self-Assessment- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    This Chief of Nuclear Safety (CNS) Report was prepared to summarize the results of the July 2013 CNS self-assessment of the Senior Technical Safety Manager Qualification Program.

  15. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward

    SciTech Connect (OSTI)

    John Collins

    2009-01-01T23:59:59.000Z

    This document presents the Next Generation Nuclear Plant (NGNP) Systems, Subsystems, and Components, establishes a baseline for the current technology readiness status, and provides a path forward to achieve increasing levels of technical maturity.

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

    SciTech Connect (OSTI)

    Larson, D.E. (ed.)

    1980-09-01T23:59:59.000Z

    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.

  17. A ThreeDimensional Finite Element Simulation for Transport of Nuclear Waste Contamination in Porous Media \\Lambda

    E-Print Network [OSTI]

    Ewing, Richard E.

    A Three­Dimensional Finite Element Simulation for Transport of Nuclear Waste Contamination for transport of nuclear­waste contamination in three­dimensional porous media are presented with a description of contamination of groundwater by high­level nuclear waste and a wide variety of other sources makes a proper

  18. Burying democracy along with UK nuclear waste Stuart Haszeldine 3 May 2012 s.haszeldine@ed.ac.uk

    E-Print Network [OSTI]

    Burying democracy along with UK nuclear waste Stuart Haszeldine 3 May 2012 s up to accept. The UK was first to deploy civil nuclear power in 1956, but waste disposal has experienced serial inertia amongst technocrats and politicians, so that problem of nuclear waste accumulating

  19. EXPERIMENTAL METHODS TO ESTIMATE ACCUMULATED SOLIDS IN NUCLEAR WASTE TANKS

    SciTech Connect (OSTI)

    Duignan, M.; Steeper, T.; Steimke, J.

    2012-12-10T23:59:59.000Z

    The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a waste tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream; Magnetic wand used to manually remove stainless steel solids from samples and the tank heel; Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas; Laser rangefinders to determine the volume and shape of the solids mounds; Core sampler to determine the stainless steel solids distribution within the solids mounds; Computer driven positioner that placed the laser rangefinders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test.

  20. Interfaces between transport and geologic disposal systems for high-level radioactive wastes and spent nuclear fuel: A new international guidance document

    SciTech Connect (OSTI)

    Pope, R.B. [Oak Ridge National Lab., TN (United States); Baekelandt, L. [Organisme National des Dechets Radioactifs et des Matieres Fissiles, Brussels (Belgium); Hoorelbeke, J.M. [CEA Agence Nationale pour la Gestion des Dechets Radioactifes (ANDRA), 75 - Paris (France); Han, K.W.; Pollog, T. [International Atomic Energy Agency, Vienna (Austria); Blackman, D. [Department of Transport, London (United Kingdom); Villagran, J.E. [Villagran Nuclear Consulting Services, Toronto, ON (Canada)

    1994-04-01T23:59:59.000Z

    An International Atomic Energy Agency (IAEA) Technical Document (TECDOC) has been developed and will be published by the IAEA. The TECDOC addresses the interfaces between the transport and geologic disposal systems for, high-level waste (HLW) and spent nuclear fuel (SNF). The document is intended to define and assist in discussing, at both the domestic and the international level, regulatory, technical, administrative, and institutional interfaces associated with HLW and SNF transport and disposal systems; it identifies and discusses the interfaces and interface requirements between the HLW and SNF, the waste transport system used for carriage of the waste to the disposal facility, and the HLW/SNF disposal facility. It provides definitions and explanations of terms; discusses systems, interfaces and interface requirements; addresses alternative strategies (single-purpose packages and multipurpose packages) and how interfaces are affected by the strategies; and provides a tabular summary of the requirements.

  1. Decommissioning and Dismantling of Liquid Waste Storage and Liquid Waste Treatment Facility from Paldiski Nuclear Site, Estonia

    SciTech Connect (OSTI)

    Varvas, M. [AS ALARA, Leetse tee 21, Paldiski, 76806 (Estonia); Putnik, H. [Delegation of the European Commission to Russia, Kadashevskaja nab. 14/1 119017 Moscow (Russian Federation); Nirvin, B.; Pettersson, S. [SKB, Box 5864, Stockholm, SE-102 40 (Sweden); Johnsson, B. [Studsvik RadWaste, Nykoping, SE-611 82 (Sweden)

    2006-07-01T23:59:59.000Z

    The Paldiski Nuclear Facility in Estonia, with two nuclear reactors was owned by the Soviet Navy and was used for training the navy personnel to operate submarine nuclear reactors. After collapse of Soviet Union the Facility was shut down and handed over to the Estonian government in 1995. In co-operation with the Paldiski International Expert Reference Group (PIERG) decommission strategy was worked out and started to implement. Conditioning of solid and liquid operational waste and dismantling of contaminated installations and buildings were among the key issues of the Strategy. Most of the liquid waste volume, remained at the Facility, was processed in the frames of an Estonian-Finnish co-operation project using a mobile wastewater purification unit NURES (IVO International OY) and water was discharged prior to the site take-over. In 1999-2002 ca 120 m{sup 3} of semi-liquid tank sediments (a mixture of ion exchange resins, sand filters, evaporator and flocculation slurry), remained after treatment of liquid waste were solidified in steel containers and stored into interim storage. The project was carried out under the Swedish - Estonian co-operation program on radiation protection and nuclear safety. Contaminated installations in buildings, used for treatment and storage of liquid waste (Liquid Waste Treatment Facility and Liquid Waste Storage) were then dismantled and the buildings demolished in 2001-2004. (authors)

  2. Identifying suitable "piercement" salt domes for nuclear waste storage sites

    SciTech Connect (OSTI)

    Kehle, R.

    1980-08-01T23:59:59.000Z

    Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes.

  3. Commercial nuclear reactors and waste: the current status

    SciTech Connect (OSTI)

    Platt, A.M.; Robinson, J.V.

    1980-04-01T23:59:59.000Z

    During the last five years, the declared size of the commercial light water reactor (LWR) nuclear power industry in the US has steadily decreased. As of January 1980, the total number of power plants had dropped to 191 from the 226 in December 31, 1974. At least another nine were cancelled in the last few months. This report was developed as the first of a series to track implications to waste management due to such changes in the declared size of the industry. For the presently declared size, key conclusions are: the declared reactors will peak at a capacity of 162 GWe and consume about 10/sup 6/ MTU as enrichment feed. As few as two repositories of about 100,000 MTHM capacity each would hold the waste. Predisposal storage (reactor basins and AFRs) would peak at less than 100,000 MTHM (in the year 2020) with one repository opening in the year 1997 and the other in the year 2020. Most of the 100,000 MTHM would have to be in AFR storage unless current practice regarding reactor basin size was radically changed.

  4. Redox reaction and foaming in nuclear waste glass melting

    SciTech Connect (OSTI)

    Ryan, J.L.

    1995-08-01T23:59:59.000Z

    This document was prepared by Pacific Northwest Laboratory (PNL) and is an attempt to analyze and estimate the effects of feed composition variables and reducing agent variables on the expected chemistry of reactions occurring in the cold cap and in the glass melt in the nuclear waste glass Slurry-fed, joule-heated melters as they might affect foaming during the glass-making process. Numerous redox reactions of waste glass components and potential feed additives, and the effects of other feed variables on these reactions are reviewed with regard to their potential effect on glass foaming. A major emphasis of this report is to examine the potential positive or negative aspects of adjusting feed with formic acid as opposed to other feed modification techniques including but not limited to use of other reducing agents. Feed modification techniques other than the use of reductants that should influence foaming behavior include control of glass melter feed pH through use of nitric acid. They also include partial replacement of sodium salts by lithium salts. This latter action (b) apparently lowers glass viscosity and raises surface tension. This replacement should decrease foaming by decreasing foam stability.

  5. Ceramic package fabrication for YMP nuclear waste disposal

    SciTech Connect (OSTI)

    Wilfinger, K.

    1994-08-01T23:59:59.000Z

    The purpose of this work is to develop alternate materials/design concepts to metal barriers for the Nevada Nuclear Waste Storage Investigations Project. There is some potential that site conditions may prove to be too aggressive for successful employment of the metal alloys under current consideration or that performance assessment models will predict metal container degradation rates that are inconsistent with the goal of substantially complete containment included in the NRC regulations. In the event that the anticipated lifetimes of metal containers are considered inadequate, alternate materials (i.e. ceramics or ceramic/metal composites) will be chosen due to superior corrosion resistance. This document was prepared using information taken from the open literature, conversations and correspondence with vendors, news releases and data presented at conferences to determine what form such a package might take. This discussion presents some ceramic material selection criteria, alternatives for the materials which might be used and alternatives for potential fabrication routes. This includes {open_quotes}stand alone{close_quotes} ceramic components and ceramic coatings/linings for metallic structures. A list of companies providing verbal or written information concerning the production of ceramic or ceramic lined waste containers appears at the end of this discussion.

  6. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    SciTech Connect (OSTI)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26T23:59:59.000Z

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  7. The Potential of Cellulosic Ethanol Production from Municipal Solid Waste: A Technical and Economic Evaluation

    E-Print Network [OSTI]

    Shi, Jian; Ebrik, Mirvat; Yang, Bin; Wyman, Charles E.

    2009-01-01T23:59:59.000Z

    e.g. waste water treatment, solid waste combustion) weretreatments. Results and Discussions Cellulosic MSW Feedstocks Municipal solid waste (

  8. The Potential of Cellulosic Ethanol Production from Municipal Solid Waste: A Technical and Economic Evaluation

    E-Print Network [OSTI]

    Shi, Jian; Ebrik, Mirvat; Yang, Bin; Wyman, Charles E.

    2009-01-01T23:59:59.000Z

    feedstocks, such as agricultural wastes and energy crops,feedstocks, such as agricultural wastes and forestry wastes,biomass, such as agricultural waste corn stover (112.7

  9. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Dole, Leslie R. [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831-6165 (United States)

    2007-07-01T23:59:59.000Z

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  10. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    SciTech Connect (OSTI)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-06-01T23:59:59.000Z

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  11. The Encapsulated Nuclear Heat Source for Proliferation-Resistant Low-Waste Nuclear Energy

    SciTech Connect (OSTI)

    Brown, N; Carelli, M; Conway, L; Dzodzo, M; Greenspan, E; Hossain, Q; Saphier, D; Shimada, H; Sienicki, J; Wade, D

    2001-04-01T23:59:59.000Z

    Encapsulated Nuclear Heat Source (ENHS) is a small innovative reactor suitable for use in developing countries. The reference design is a SOMWe lead-bismuth eutectic (Pb-Bi) cooled fast reactor. It is designed so that the fuel is installed and sealed into the reactor module at the factory. The nuclear controls, a major portion of the instrumentation and the Pb-Bi covering the core are also installed at the factory. At the site of operations the reactor module is inserted into a pool of Pb-Bi that contains the steam generators. Major components, such as the pool vessel and steam generators, are permanent and remain in place while the reactor module is replaced every 15 years. At the end of life the sealed reactor module is removed and returned to an internationally controlled recycling center. Thus, the ENHS provides a unique capability for ensuring the security of the nuclear fuel throughout its life. The design also can minimize the user country investment in nuclear technology and staff. Following operation and return of the module to the recycling facility, the useable components, including the fuel, are refurbished and available for reuse. A fuel cycle compatible with this approach has been identified that reduces the amount of nuclear waste.

  12. Proceedings of the symposium on Scientific Basis for Nuclear Waste Management XXX

    SciTech Connect (OSTI)

    Dunn, Darrell [ed. Southwest Research Inst., San Antonio, Texas (United States); Poinssot, Christophe [ed. CEA-Saclay, 91191 Gif-sur-Yvette cedex (France); Begg, Bruce [ed. Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia)

    2007-07-01T23:59:59.000Z

    Management of nuclear waste remains an important international topic that includes reprocessing of commercial nuclear fuel, waste-form design and development, storage and disposal packaging, the process of repository site selection, system design, and performance assessment. Requirements to manage and dispose of materials from the production of nuclear weapons, and the renewed interest in nuclear power, in particular through the Generation IV Forum and the Advanced Fuel Cycle Initiative, can be expected to increase the need for scientific advances in waste management. A broad range of scientific and engineering disciplines is necessary to provide safe and effective solutions and address complex issues. This volume offers an interdisciplinary perspective on materials-related issues associated with nuclear waste management programs. Invited and contributed papers cover a wide range of topics including studies on: spent fuel; performance assessment and models; waste forms for low- and intermediate-level waste; ceramic and glass waste forms for plutonium and high-level waste; radionuclides; containers and engineered barriers; disposal environments and site characteristics; and partitioning and transmutation.

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

    DOE Patents [OSTI]

    Boatner, Lynn A. (Oak Ridge, TN); Sales, Brian C. (Oak Ridge, TN)

    1989-01-01T23:59:59.000Z

    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.

  14. Vanadium and Chromium Redox Behavior in borosilicate Nuclear Waste Glasses

    SciTech Connect (OSTI)

    D McKeown; I Muller; H Gan; Z Feng; C Viragh; I Pegg

    2011-12-31T23:59:59.000Z

    X-ray absorption spectroscopy (XAS) was used to characterize vanadium (V) and chromium (Cr) environments in low activity nuclear waste (LAW) glasses synthesized under a variety of redox conditions. V{sub 2}O{sub 5} was added to the melt to improve sulfur incorporation from the waste; however, at sufficiently high concentrations, V increased melt foaming, which lowered melt processing rates. Foaming may be reduced by varying the redox conditions of the melt, while small amounts of Cr are added to reduce melter refractory corrosion. Three parent glasses were studied, where CO-CO{sub 2} mixtures were bubbled through the corresponding melt for increasing time intervals so that a series of redox-adjusted-glasses was synthesized from each parent glass. XAS data indicated that V and Cr behaviors are significantly different in these glasses with respect to the cumulative gas bubbling times: V{sup 4+}/V{sub total} ranges from 8 to 35%, while Cr{sup 3+}/Cr{sub total} can range from 15 to 100% and even to population distributions including Cr{sup 2+}. As Na-content decreased, V, and especially, Cr became more reduced, when comparing equivalent glasses within a series. The Na-poor glass series show possible redox coupling between V and Cr, where V{sup 4+} populations increase after initial bubbling, but as bubbling time increases, V{sup 4+} populations drop to near the level of the parent glass, while Cr becomes more reduced to the point of having increasing Cr{sup 2+} populations.

  15. Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

    SciTech Connect (OSTI)

    French, Sean B. [Los Alamos National Laboratory; Shuman, Robert [WPS: WASTE PROJECTS AND SERVICES

    2012-04-17T23:59:59.000Z

    The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D and D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare and maintain site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on-site and off-site exposure scenarios. The assessments are based on existing site and disposal facility data and on assumptions about future rates and methods of waste disposal. The accuracy of the performance assessment and composite analysis depends upon the validity of the data used and assumptions made in conducting the analyses. If changes in these data and assumptions are significant, they may invalidate or call into question certain aspects of the analyses. For example, if the volumes and activities of waste disposed of during the remainder of the disposal facility's lifetime differ significantly from those projected, the doses projected by the analyses may no longer apply. DOE field sites are required to implement a performance assessment and composite analysis maintenance program. The purpose of this program is to ensure the continued applicability of the analyses through incremental improvement of the level of understanding of the disposal site and facility. Site personnel are required to conduct field and experimental work to reduce the uncertainty in the data and models used in the assessments. Furthermore, they are required to conduct periodic reviews of waste receipts, comparing them to projected waste disposal rates. The radiological inventory for Area G was updated in conjunction with Revision 4 of the performance assessment and composite analysis (Shuman, 2008). That effort used disposal records and other sources of information to estimate the quantities of radioactive waste that have been disposed of at Area G from 1959, the year the facility started receiving waste on a routine basis, through 2007. It also estimated the quantities of LLW that will require disposal from 2008 through 2044, the year in which it is assumed that disposal operations at Area G will cease. This report documents the fourth review of Area G disposal receipts since the inventory was updated and examines information for waste placed in the ground during fiscal years (FY) 2008 through 2011. The primary objective of the disposal receipt review is to ensure that the future waste inventory projections developed for the performance assessment and composite analysis are consistent with the actual types and quantities of waste being disposed of at Area G. Toward this end, the disposal data that are the subject of this review are used to update the future waste inventory projections for the disposal facility. These projections are compared to the future inventory projections that were develope

  16. GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions

    Broader source: Energy.gov [DOE]

    Consistent with the Administration's commitment to transparency, DOE General Counsel Scott Blake Harris has decided that all future determinations as to the adequacy of the Nuclear Waste Fund fee...

  17. An evaluation of the feasibility of disposal of nuclear waste in very deep boreholes

    E-Print Network [OSTI]

    Anderson, Victoria Katherine, 1980-

    2004-01-01T23:59:59.000Z

    Deep boreholes, 3 to 5 km into igneous rock, such as granite, are evaluated for next- generation repository use in the disposal of spent nuclear fuel and other high level waste. The primary focus is on the stability and ...

  18. Feasibility of lateral emplacement in very deep borehole disposal of high level nuclear waste

    E-Print Network [OSTI]

    Gibbs, Jonathan Sutton

    2010-01-01T23:59:59.000Z

    The U.S. Department of Energy recently filed a motion to withdraw the Nuclear Regulatory Commission license application for the High Level Waste Repository at Yucca Mountain in Nevada. As the U.S. has focused exclusively ...

  19. Effective thermal conductivity measurements relevant to deep borehole nuclear waste disposal

    E-Print Network [OSTI]

    Shaikh, Samina

    2007-01-01T23:59:59.000Z

    The objective of this work was to measure the effective thermal conductivity of a number of materials (particle beds, and fluids) proposed for use in and around canisters for disposal of high level nuclear waste in deep ...

  20. Feasibility of very deep borehole disposal of US nuclear defense wastes

    E-Print Network [OSTI]

    Dozier, Frances Elizabeth

    2011-01-01T23:59:59.000Z

    This thesis analyzes the feasibility of emplacing DOE-owned defense nuclear waste from weapons production into a permanent borehole repository drilled ~4 km into granite basement rock. Two canister options were analyzed ...

  1. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-06-17T23:59:59.000Z

    This document was developed to support the documented safety analysis (DSA) and describes the process and basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. Revision 1 incorporates ORP IRT comments to enhance the technical presentation and also makes editorial changes. This technical basis document was developed to support the documented safety analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence.

  2. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GRIGSBY, J.M.

    2005-03-03T23:59:59.000Z

    Technical Basis Document for the Nuclear Criticality Representative Accident and Associate Represented Hazardous Conditions. Revision 2 of RPP-12371 provides accident consequence estimates for a hypothetical criticality event in an above grade facility (e.g. DBVS, CH-TRUM, and S-109 PWRS). This technical basis document was developed to support RPP-13033, ''Tank Farms Documented Safety Analysis (DSA)'', and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  3. Engineered barrier development for a nuclear waste repository in basalt: an integration of current knowledge

    SciTech Connect (OSTI)

    Smith, M.J.

    1980-05-01T23:59:59.000Z

    This document represents a compilation of data and interpretive studies conducted as part of the engineered barriers program of the Basalt Waste Isolation Project. The overall objective of these studies is to provide information on barrier system designs, emplacement and isolation techniques, and chemical reactions expected in a nuclear waste repository located in the basalts underlying the Hanford Site within the state of Washington. Backfills, waste-basalt interactions, sorption, borehole plugging, etc., are among the topics discussed.

  4. Precipitation process for the removal of technetium values from nuclear waste solutions

    DOE Patents [OSTI]

    Walker, D.D.; Ebra, M.A.

    1985-11-21T23:59:59.000Z

    High efficiency removal of techetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  5. Revised rail-stop exposure model for incident-free transport of nuclear waste

    SciTech Connect (OSTI)

    Ostmeyer, R.M.

    1986-02-01T23:59:59.000Z

    This report documents a model for estimating railstop doses that occur during incident-free transport of nuclear waste by rail. The model, which has been incorporated into the RADTRAN III risk assessment code, can be applied to general freight and dedicated train shipments of waste.

  6. Technical progress report for the quarter 1 October-31 December 1980

    SciTech Connect (OSTI)

    Not Available

    1981-01-01T23:59:59.000Z

    This report describes the technical accomplishments on the commercial nuclear waste management programs and on the geologic disposal of nuclear wastes. The program is organized into eight tasks: systems, waste package, site, repository, regulatory and institutional, test facilities and excavations, land acquisition, and program management. (DLC)

  7. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    released by the buried wastes and heat remain­ ing in theOF 10-YEAR-OLD WASTES Waste Heat Source C h a r a c t e r ia t e r s e c t i o n s . WASTE HEAT SOURCE CHARACTERIZATION

  8. Nuclear Safety. Technical Progress Journal, October--December 1991: Volume 32, No. 4

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  9. Hydrogeologic effects of natural disruptive events on nuclear waste repositories

    SciTech Connect (OSTI)

    Davis, S.N.

    1980-06-01T23:59:59.000Z

    Some possible hydrogeologic effects of disruptive events that may affect repositories for nuclear wastte are described. A very large number of combinations of natural events can be imagined, but only those events which are judged to be most probable are covered. Waste-induced effects are not considered. The disruptive events discussed above are placed into four geologic settings. Although the geology is not specific to given repository sites that have been considered by other agencies, the geology has been generalized from actual field data and is, therefore, considered to be physically reasonable. The geologic settings considered are: (1) interior salt domes of the Gulf Coast, (2) bedded salt of southeastern New Mexico, (3) argillaceous rocks of southern Nevanda, and (4) granitic stocks of the Basin and Range Province. Log-normal distributions of permeabilities of rock units are given for each region. Chapters are devoted to: poresity and permeability of natural materials, regional flow patterns, disruptive events (faulting, dissolution of rock forming minerals, fracturing from various causes, rapid changes of hydraulic regimen); possible hydrologic effects of disruptive events; and hydraulic fracturing.

  10. Helium solubility in SON68 nuclear waste glass

    SciTech Connect (OSTI)

    Fares, Toby; Peuget, Sylvain; Bouty, Olivier; Broudic, Veronique; Maugeri, Emilio; Bes, Rene; Jegou, Christophe [CEA, DEN, DTCD SECM LMPA, F-30207 Marcoule, Bagnols Sur Cez, (France); Chamssedine, Fadel; Sauvage, Thierry [CNRS, CEMHTI, F-245071 Orleans, (France); Deschanels, Xavier [LNAR, Marcoule Inst Separat Chem, F-30207 Bagnols Sur Ceze, (France)

    2012-12-15T23:59:59.000Z

    Helium behavior in a sodium borosilicate glass (SON68) dedicated to the immobilization of high-level nuclear waste is examined. Two experimental approaches on nonradioactive glass specimens are implemented: pressurized helium infusion experiments and {sup 3}He ion implantation experiments. The temperature variation of helium solubility in SON68 glass was determined and analyzed with the harmonic oscillator model to determine values of the energy of interaction E(0) at the host sites (about -4000 J/mol), the vibration frequency (about 1.7 x 10{sup 11} s{sup -1}), and the density of solubility sites (2.2 x 10{sup 21} sites cm{sup -3}). The implantation experiments show that a non diffusive transport phenomenon (i.e., athermal diffusion) is involved in the material when the helium concentration exceeds 2.3 x 10{sup 21} He cm{sup -3}, and thus probably as soon as it exceeds the density of solubility sites accessible to helium in the glass. We propose that this transport mechanism could be associated with the relaxation of the stress gradient induced by the implanted helium profile, which is favored by the glass damage. Microstructural characterization by TEM and ESEM of glass specimens implanted with high helium concentrations showed a homogeneous microstructure free of bubbles, pores, or cracking at a scale of 10 nm. (authors)

  11. Guidance document for revision of DOE Order 5820.2A, Radioactive Waste Technical Support Program. Revision 1

    SciTech Connect (OSTI)

    Kudera, D.E.; McMurtrey, C.D.; Meagher, B.G.

    1993-04-01T23:59:59.000Z

    This document provides guidance for the revision of DOE Order 5820.2A, ``Radioactive Waste Management.`` Technical Working Groups have been established and are responsible for writing the revised order. The Technical Working Groups will use this document as a reference for polices and procedures that have been established for the revision process. The overall intent of this guidance is to outline how the order will be revised and how the revision process will be managed. In addition, this document outlines technical issues considered for inclusion by a Department of Energy Steering Committee.

  12. The Potential of Cellulosic Ethanol Production from Municipal Solid Waste: A Technical and Economic Evaluation

    E-Print Network [OSTI]

    Shi, Jian; Ebrik, Mirvat; Yang, Bin; Wyman, Charles E.

    2009-01-01T23:59:59.000Z

    1996 19950414. Municipal solid waste processing facility andconversion of municipal-solid-waste to ethanol. Biotechnol.Bioconversion of municipal solid waste to glucose for bio-

  13. Na, Mg, Ni and Cs distribution and speciation after long-term alteration of a simulated nuclear waste glass

    E-Print Network [OSTI]

    distribution and speciation of Na, Mg, Ni and Cs in a simulated (inactive) nuclear waste glass were studied and Cs represent dose determining long-lived radionuclides (59 Ni, 135 Cs) in vitrified nuclear wasteNa, Mg, Ni and Cs distribution and speciation after long-term alteration of a simulated nuclear

  14. Technical Basis for U. S. Department of Energy Nuclear Safety Policy, DOE Policy 420.1

    Broader source: Energy.gov [DOE]

    This document provides the technical basis for the Department of Energy (DOE) Policy (P) 420.1, Nuclear Safety Policy, dated 2-8-2011. It includes an analysis of the revised Policy to determine whether it provides the necessary and sufficient high-level expectations that will lead DOE to establish and implement appropriate requirements to assure protection of the public, workers, and the environment from the hazards of DOE’s operation of nuclear facilities.

  15. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect (OSTI)

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi [Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Kawashima, Masatoshi [Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kotake, Shoji [The Japan Atomic Power Company, 1-1, Kanda-Mitoshiro-cho, Chiyoda-ku, Tokyo 101-0053 (Japan); Fuji-ie, Yoichi [Nuclear Salon Fuji-ie, 1-11-10, Yushima, Bunkyo-ku, Tokyo 113-0034 (Japan)

    2013-07-01T23:59:59.000Z

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  16. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

    2000-11-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  17. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    SciTech Connect (OSTI)

    Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

    2000-10-31T23:59:59.000Z

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  18. Performance evaluation of the technical capabilities of DOE sites for disposal of mixed low-level waste. Volume 2: Technical basis and discussion of results

    SciTech Connect (OSTI)

    Waters, R.D.; Gruebel, M.M.; Hospelhorn, M.B. [and others

    1996-03-01T23:59:59.000Z

    A team of analysts designed and conducted a performance evaluation to estimate the technical capabilities of fifteen Department of Energy sites for disposal of mixed low-level waste (i.e., waste that contains both low-level radioactive materials and hazardous constituents). Volume 1 summarizes the process for selecting the fifteen sites, the methodology used in the evaluation, and the conclusions derived from the evaluation. Volume 2 first describes the screening process used to determine the sites to be considered in the PEs. This volume then provides the technical details of the methodology for conducting the performance evaluations. It also provides a comparison and analysis of the overall results for all sites that were evaluated. Volume 3 contains detailed evaluations of the fifteen sites and discussions of the results for each site.

  19. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    SciTech Connect (OSTI)

    NA

    2005-10-25T23:59:59.000Z

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

  20. Processing constraints on high-level nuclear waste glasses for Hanford Waste Vitrification Plant

    SciTech Connect (OSTI)

    Hrma, P. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-31T23:59:59.000Z

    The work presented in this paper is a part of a major technology program supported by the US Department of Energy (DOE) in preparation for the planned operation of the Hanford Waste Vitrification Plant (HWVP). Because composition of Hanford waste varies greatly, processability is a major concern for successful vitrification. This paper briefly surveys general aspects of waste glass processability and then discusses their ramifications for specific examples of Hanford waste streams.

  1. TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

    SciTech Connect (OSTI)

    VAIL, T.S.

    2007-03-30T23:59:59.000Z

    This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously un-overpacked package to subsequently require overpacking. Therefore, the polyurea coating can not be credited to avoid the need to overpack a package or enhance the transportation safety of a structurally sound package that has polyurea on the exterior.

  2. alkaline nuclear wastes: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Geosciences Websites Summary: -solid waste for CO2 mitigation and reduction of greenhouse effect gases into the atmosphere. ? 2008 ElsevierCarbonation of alkaline paper mill...

  3. alkaline nuclear waste: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Geosciences Websites Summary: -solid waste for CO2 mitigation and reduction of greenhouse effect gases into the atmosphere. ? 2008 ElsevierCarbonation of alkaline paper mill...

  4. Technical basis for the nuclear criticality representative accident and associated represented hazardous conditions

    SciTech Connect (OSTI)

    CARSON, D.M.

    2003-03-20T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  5. The development of a management strategy for interim storage and final disposal of nuclear wastes

    SciTech Connect (OSTI)

    Engelmann, H.J.; Popp, F.W. [Deutsche Gesellschaft zum Bau and Betrieb von Endglagern fuer Abfallostofe mbH, Peine (Germany); Arntzen, P.; Botzem, W. [NUKEM GmbH, Alzenau (Germany); Soucek, B. [Czech Power Board, Prague (Czech Republic)

    1993-12-31T23:59:59.000Z

    The overall waste management strategy for alternative routes from reactor to final disposal, including dry interim storage, is discussed. Within the framework of a preliminary structure plan possible technical solutions must be investigated, and with sufficient relevant information available the future progress of the project, can be addressed on the base of a decision analysis.

  6. Can we talk? Communications management for the Waste Isolation Pilot Plant, a complex nuclear waste management project

    SciTech Connect (OSTI)

    Goldstein, S.A.; Pullen, G.M.; Brewer, D.R.

    1995-07-01T23:59:59.000Z

    Sandia Nuclear Waste Management Program is pursuing for DOE an option for permanently disposing radioactive waste in deep geologic repositories. Included in the Program are the Waste Isolation Pilot Plant (WIPP) Project for US defense program mixed waste the Yucca Mountain Project (YMP) for spent power reactor fuel and vitrified high-level waste, projects for other waste types, and development efforts in environmental decision support technologies. WIPP and YMP are in the public arena, of a controversial nature, and provide significant management challenges. Both projects have large project teams, multiple organization participants, large budgets, long durations, are very complex, have a high degree of programmatic risk, and operate in an extremely regulated environment requiring legal defensibility. For environmental projects like these to succeed, SNL`s Program is utilizing nearly all areas in PMI`s Project Management Body of Knowledge (PMBOK) to manage along multiple project dimensions such as the physical sciences (e.g., geophysics and geochemistry; performance assessment; decision analysis) management sciences (controlling the triple constraint of performance, cost and schedule), and social sciences (belief systems; public participation; institutional politics). This discussion focuses primarily on communication challenges active on WIPP. How is the WIPP team meeting the challenges of managing communications?`` and ``How are you approaching similar challenges?`` will be questions for a dialog with the audience.

  7. A report on high-level nuclear waste transportation: Prepared pursuant to assembly concurrent resolution No. 8 of the 1987 Nevada Legislature

    SciTech Connect (OSTI)

    NONE

    1988-12-01T23:59:59.000Z

    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.

  8. Computational Efficient Upscaling Methodology for Predicting Thermal Conductivity of Nuclear Waste forms

    SciTech Connect (OSTI)

    Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

    2011-09-28T23:59:59.000Z

    This study evaluated different upscaling methods to predict thermal conductivity in loaded nuclear waste form, a heterogeneous material system. The efficiency and accuracy of these methods were compared. Thermal conductivity in loaded nuclear waste form is an important property specific to scientific researchers, in waste form Integrated performance and safety code (IPSC). The effective thermal conductivity obtained from microstructure information and local thermal conductivity of different components is critical in predicting the life and performance of waste form during storage. How the heat generated during storage is directly related to thermal conductivity, which in turn determining the mechanical deformation behavior, corrosion resistance and aging performance. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling models were developed and implemented. Due to the absence of experimental data, prediction results from finite element method (FEM) were used as reference to determine the accuracy of different upscaling models. Micrographs from different loading of nuclear waste were used in the prediction of thermal conductivity. Prediction results demonstrated that in term of efficiency, boundary models (Taylor and Sachs model) are better than self consistent model, statistical upscaling method and FEM. Balancing the computation resource and accuracy, statistical upscaling is a computational efficient method in predicting effective thermal conductivity for nuclear waste form.

  9. Review of Used Nuclear Fuel Storage and Transportation Technical Gap

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin ofEnergy at Waste-to-Energy usingofRetrofitting DoorsReview of SAR

  10. Review of Used Nuclear Fuel Storage and Transportation Technical Gap

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin ofEnergy at Waste-to-Energy usingofRetrofitting DoorsReview of SARAnalysis |

  11. Nuclear-waste-management. Quarterly progress report, July-September 1981

    SciTech Connect (OSTI)

    Chikalla, T.D.; Powell, J.A. (comps.)

    1981-12-01T23:59:59.000Z

    Progress reports and summaries are presented for the following: high-level waste process development, alternate waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and fuel pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

  12. River Protection Project (RPP) Tank Waste Retrieval and Disposal Mission Technical Baseline Summary Description

    SciTech Connect (OSTI)

    DOVALLE, O.R.

    1999-12-29T23:59:59.000Z

    This document is one of the several documents prepared by Lockheed Martin Hanford Corp. to support the U. S. Department of Energy's Tank Waste Retrieval and Disposal mission at Hanford. The Tank Waste Retrieval and Disposal mission includes the programs necessary to support tank waste retrieval; waste feed, delivery, storage, and disposal of immobilized waste; and closure of the tank farms.

  13. State of the art review of radioactive waste volume reduction techniques for commercial nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1980-04-01T23:59:59.000Z

    A review is made of the state of the art of volume reduction techniques for low level liquid and solid radioactive wastes produced as a result of: (1) operation of commercial nuclear power plants, (2) storage of spent fuel in away-from-reactor facilities, and (3) decontamination/decommissioning of commercial nuclear power plants. The types of wastes and their chemical, physical, and radiological characteristics are identified. Methods used by industry for processing radioactive wastes are reviewed and compared to the new techniques for processing and reducing the volume of radioactive wastes. A detailed system description and report on operating experiences follow for each of the new volume reduction techniques. In addition, descriptions of volume reduction methods presently under development are provided. The Appendix records data collected during site surveys of vendor facilities and operating power plants. A Bibliography is provided for each of the various volume reduction techniques discussed in the report.

  14. Numerical study of the thm effects on the near-field safety of a hypothetical nuclear waste repository - bmt1 of the decovalex iii project. part 1: conceptualization and characterization of the problems and summary of results

    E-Print Network [OSTI]

    2004-01-01T23:59:59.000Z

    T-H-M Processes for Nuclear Waste Repositories. Executivesafety of a hypothetical nuclear waste repository – BMT1 ofof a hypothetical nuclear waste repository at the near-fiel

  15. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1997-03-01T23:59:59.000Z

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

  16. The dilemma of siting a high-level nuclear waste repository

    SciTech Connect (OSTI)

    Easterline, D.; Kunreuther, H.

    1995-12-31T23:59:59.000Z

    This books presents a siting process that the authors believe will prove successful within the adversarial world that characterizes most attempts to build waste-disposal facilities. They come to the following conclusions: a volunatary siting process stands the best chance of breaking the `not-in-my-backyard` problem; and without public acknowledgement that a facility is needed, any proposal to build a high-level nuclear waste storage facility will meet with opposition.

  17. Nuclear waste vitrification: electric melting and glass formulation

    SciTech Connect (OSTI)

    Hrma, Pavel R.

    2007-07-10T23:59:59.000Z

    The Hanford Site contains 177 underground tanks with radioactive waste that will be vitrified, i.e., immobilized by converting it to glass in electric melters. After pretreatment, the waste slurry will be mixed with glass-forming minerals, and the resulting feed will be charged into the melter. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical properties that guarantee that the glass is easily made and resists environmental degradation. On heating, the feed undergoes complex reactions. The large variability of waste compositions presents numerous technological challenges: undesirable insoluble solids and molten salts may segregate; foam may hinder heat transfer and slows down the process; and on cooling, the glass may precipitate crystalline phases.

  18. Technical Note: Evaluation of Effective Microorganisms (EM) In Solid Waste Management

    E-Print Network [OSTI]

    Sekeran, V.; Balaji, C.; Bhagavathipushpa, T.

    2005-01-01T23:59:59.000Z

    Microorganisms (EM) In Solid Waste Management V. Sekeran C.is the treatment, disposal and/or recycling of solid wastes.Generally solid waste from a municipality consists of

  19. Solid Waste Integrated Forecast Technical (SWIFT) Report FY2001 to FY2046 Volume 1

    SciTech Connect (OSTI)

    BARCOT, R.A.

    2000-08-31T23:59:59.000Z

    This report provides up-to-date life cycle information about the radioactive solid waste expected to be managed by Hanford's Waste Management (WM) Project from onsite and offsite generators. It includes: an overview of Hanford-wide solid waste to be managed by the WM Project; program-level and waste class-specific estimates; background information on waste sources; and comparisons to previous forecasts and other national data sources. This report does not include: waste to be managed by the Environmental Restoration (EM-40) contractor (i.e., waste that will be disposed of at the Environmental Restoration Disposal Facility (ERDF)); waste that has been received by the WM Project to date (i.e., inventory waste); mixed low-level waste that will be processed and disposed by the River Protection Program; and liquid waste (current or future generation). Although this report currently does not include liquid wastes, they may be added as information becomes available.

  20. Use of tangential filtration unit for processing liquid waste from nuclear laundries

    SciTech Connect (OSTI)

    Augustin, X.; Buzonniere, A. de [Technicatome, Gif-sur-Yvette (France); Barnier, H. [CEA Cadarache, St. Paul-lez-Durance (France)

    1993-12-31T23:59:59.000Z

    Nuclear facilities produce large quantities of weakly contaminated effluents charged with insoluble and soluble products. In collaboration with CEA, TECHNICATOME has developed an ultrafiltration process for liquid waste from nuclear facilities associated with prior insolubilization of radiochemical activity. This process, seeded ultrafiltration, is based on the use of a decloggable mineral filter media and combines very high separation efficiency with long membrane life. The efficiency of the tangential filtration unit which has been processing effluents from the Cadarache Nuclear Research Center (CEA-France) nuclear facility since mid-1988, has been confirmed on several sites.

  1. Radioactive Waste Management and Nuclear Facility Decommissioning Progress in Iraq - 13216

    SciTech Connect (OSTI)

    Al-Musawi, Fouad; Shamsaldin, Emad S.; Jasim, Hadi [Ministry of Science and Technology (MoST), Al-Jadraya, P.O. Box 0765, Baghdad (Iraq)] [Ministry of Science and Technology (MoST), Al-Jadraya, P.O. Box 0765, Baghdad (Iraq); Cochran, John R. [Sandia National Laboratories1, New Mexico, Albuquerque New Mexico 87185 (United States)] [Sandia National Laboratories1, New Mexico, Albuquerque New Mexico 87185 (United States)

    2013-07-01T23:59:59.000Z

    Management of Iraq's radioactive wastes and decommissioning of Iraq's former nuclear facilities are the responsibility of Iraq's Ministry of Science and Technology (MoST). The majority of Iraq's former nuclear facilities are in the Al-Tuwaitha Nuclear Research Center located a few kilometers from the edge of Baghdad. These facilities include bombed and partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. Within these facilities are large numbers of silos, approximately 30 process or waste storage tanks and thousands of drums of uncharacterised radioactive waste. There are also former nuclear facilities/sites that are outside of Al-Tuwaitha and these include the former uranium processing and waste storage facility at Jesira, the dump site near Adaya, the former centrifuge facility at Rashdiya and the former enrichment plant at Tarmiya. In 2005, Iraq lacked the infrastructure needed to decommission its nuclear facilities and manage its radioactive wastes. The lack of infrastructure included: (1) the lack of an organization responsible for decommissioning and radioactive waste management, (2) the lack of a storage facility for radioactive wastes, (3) the lack of professionals with experience in decommissioning and modern waste management practices, (4) the lack of laws and regulations governing decommissioning or radioactive waste management, (5) ongoing security concerns, and (6) limited availability of electricity and internet. Since its creation eight years ago, the MoST has worked with the international community and developed an organizational structure, trained staff, and made great progress in managing radioactive wastes and decommissioning Iraq's former nuclear facilities. This progress has been made, despite the very difficult implementing conditions in Iraq. Within MoST, the Radioactive Waste Treatment and Management Directorate (RWTMD) is responsible for waste management and the Iraqi Decommissioning Directorate (IDD) is responsible for decommissioning activities. The IDD and the RWTMD work together on decommissioning projects. The IDD has developed plans and has completed decommissioning of the GeoPilot Facility in Baghdad and the Active Metallurgical Testing Laboratory (LAMA) in Al-Tuwaitha. Given this experience, the IDD has initiated work on more dangerous facilities. Plans are being developed to characterize, decontaminate and decommission the Tamuz II Research Reactor. The Tammuz Reactor was destroyed by an Israeli air-strike in 1981 and the Tammuz II Reactor was destroyed during the First Gulf War in 1991. In addition to being responsible for managing the decommissioning wastes, the RWTMD is responsible for more than 950 disused sealed radioactive sources, contaminated debris from the first Gulf War and (approximately 900 tons) of naturally-occurring radioactive materials wastes from oil production in Iraq. The RWTMD has trained staff, rehabilitated the Building 39 Radioactive Waste Storage building, rehabilitated portions of the French-built Radioactive Waste Treatment Station, organized and secured thousands of drums of radioactive waste organized and secured the stores of disused sealed radioactive sources. Currently, the IDD and the RWTMD are finalizing plans for the decommissioning of the Tammuz II Research Reactor. (authors)

  2. Glass former composition and method for immobilizing nuclear waste using the same

    DOE Patents [OSTI]

    Cadoff, Laurence H. (Wilkins Township, Allegheny County, PA); Smith-Magowan, David B. (Washington, DC)

    1988-01-01T23:59:59.000Z

    An alkoxide glass former composition has silica-containing constituents present as solid particulates of a particle size of 0.1 to 0.7 micrometers in diameter in a liquid carrier phase substantially free of dissolved silica. The glass former slurry is resistant to coagulation and may contain other glass former metal constituents. The immobilization of nuclear waste employs the described glass former by heating the same to reduce the volume, mixing the same with the waste, and melting the resultant mixture to encapsulate the waste in the resultant glass.

  3. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, J.P.; Marek, J.C.

    1987-02-25T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper(II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the original organic compounds, is subsequently blended with high level radioactive sludge land transferred to a vitrification facility for processing into borosilicate glass for long-term storage. 2 figs., 3 tabs.

  4. Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries

    DOE Patents [OSTI]

    Doherty, Joseph P. (Elkton, MD); Marek, James C. (Augusta, GA)

    1989-01-01T23:59:59.000Z

    A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper (II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the orginal organic compounds, is subsequently blended with high level radioactive sludge and transferred to a virtrification facility for processing into borosilicate glass for long-term storage.

  5. EA-1962: Analysis for Below Grade Suspect Transuranic (TRU) Waste at Technical Area (TA)-54

    Broader source: Energy.gov [DOE]

    DOE is preparing an EA to evaluate the legacy suspect transuranic (TRU) waste at Area G for the purposes of reclassification of waste type and determination of a final disposal path. Per DOE Order 435.1, Change 1, Radioactive Waste Management, and its associated guide, legacy waste at Los Alamos National Laboratory that contained TRU waste was stored and managed as TRU waste. The waste was given an interim classification for the purposes of applying the most restrictive standard until the waste could be adequately characterized and a final determination on the disposition classification was made.

  6. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    SciTech Connect (OSTI)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I. [Los Alamos Technical Associates, Inc., NM (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-10-01T23:59:59.000Z

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

  7. Record of Technical Change for Corrective Action Plan for Corrective Action Unit 140: Waste Dumps, burn Pits, and Storage Area, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    U.S. Department of Energy, National Nuclear Security Administration, Nevada Site Office; Bechtel Nevada

    2005-01-05T23:59:59.000Z

    Record of Technical Change for Corrective Action Plan for Corrective Action Unit 140: Waste Dumps, Burn Pits, and Storage Area, Nevada Test Site, Nevada (DOE/NV--963-Rev 2, dated November 2004).

  8. NEAMS Nuclear Waste Management IPSC : evaluation and selection of tools for the quality environment.

    SciTech Connect (OSTI)

    Bouchard, Julie F.; Stubblefield, William Anthony; Vigil, Dena M.; Edwards, Harold Carter (Org. 1444 : Multiphysics Simulation Technology)

    2011-05-01T23:59:59.000Z

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. These M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. This report documents an evaluation of the needs, options, and tools selected for the NEAMS Nuclear Waste Management IPSC quality environment. The objective of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management IPSC) program element is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to assess quantitatively the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. This objective will be fulfilled by acquiring and developing M&S capabilities, and establishing a defensible level of confidence in these M&S capabilities. The foundation for assessing the level of confidence is based upon the rigor and results from verification, validation, and uncertainty quantification (V&V and UQ) activities. M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. NEAMS Nuclear Waste Management IPSC V&V and UQ practices and evidence management goals are documented in the V&V Plan. This V&V plan includes a description of the quality environment into which M&S capabilities are imported and V&V and UQ activities are managed. The first phase of implementing the V&V plan is to deploy an initial quality environment through the acquisition and integration of a set of software tools. An evaluation of the needs, options, and tools selected for the quality environment is given in this report.

  9. Fitness for duty in the nuclear industry: Update of the technical issues 1996

    SciTech Connect (OSTI)

    Durbin, N.; Grant, T. [eds.] [Battelle Seattle Research Center, WA (United States)

    1996-05-01T23:59:59.000Z

    The purpose of this report is to provide an update of information on the technical issues surrounding the creation, implementation, and maintenance of fitness-for-duty (FFD) policies and programs. It has been prepared as a resource for Nuclear Regulatory Commission (NRC) and nuclear power plant personnel who deal with FFD programs. It contains a general overview and update on the technical issues that the NRC considered prior to the publication of its original FFD rule and the revisions to that rule (presented in earlier NUREG/CRs). It also includes chapters that address issues about which there is growing concern and/or about which there have been substantial changes since NUREG/CR-5784 was published. Although this report is intended to support the NRC`s rule making on fitness for duty, the conclusions of the authors of this report are their own and do not necessarily represent the opinions of the NRC.

  10. A compilation of reports of the Advisory Committee on nuclear waste, July 1996--June 1997

    SciTech Connect (OSTI)

    NONE

    1997-08-01T23:59:59.000Z

    This compilation contains 11 reports issued by the Advisory Committee on Nuclear Waste (ACNW) during the ninth year of its operation. The reports were submitted to the Chairman and Commissioners of the U.S. Nuclear Regulatory Commission. All reports prepared by the Committee have been made available to the public through the NRC Public Document Room, the U.S. Library of Congress, and the internet at http://www.nrc.gov/ACRSACNW.

  11. Reproduced with permission of the copyright owner. Further reproduction prohibited without permission. Overcoming tunnel vision: Redirecting the U.S. high-level nuclear waste program

    E-Print Network [OSTI]

    Kammen, Daniel M.

    permission. Overcoming tunnel vision: Redirecting the U.S. high-level nuclear waste program James Flynn

  12. Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

    2013-10-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

  13. Management of salt waste from electrochemical processing of used nuclear fuel

    SciTech Connect (OSTI)

    Simpson, M.F.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415 (United States); Lee, J.; Wang, Y. [Sandia National Laboratory, Albuquerque, NM (United States); Versey, J.; Phongikaroon, S. [University of Idaho, Idaho Falls, ID (United States)

    2013-07-01T23:59:59.000Z

    Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

  14. Technical cooperation on nuclear security between the United States and China : review of the past and opportunities for the future.

    SciTech Connect (OSTI)

    Pregenzer, Arian Leigh

    2011-12-01T23:59:59.000Z

    The United States and China are committed to cooperation to address the challenges of the next century. Technical cooperation, building on a long tradition of technical exchange between the two countries, can play an important role. This paper focuses on technical cooperation between the United States and China in the areas of nonproliferation, arms control and other nuclear security topics. It reviews cooperation during the 1990s on nonproliferation and arms control under the U.S.-China Arms Control Exchange, discusses examples of ongoing activities under the Peaceful Uses of Technology Agreement to enhance security of nuclear and radiological material, and suggests opportunities for expanding technical cooperation between the defense nuclear laboratories of both countries to address a broader range of nuclear security topics.

  15. THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED BY MO K-EDGE X-RAY ABSORPTION

    E-Print Network [OSTI]

    Sheffield, University of

    THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED of molybdenum in model UK high level nuclear waste glasses was investigated by X-ray Absorption Spectroscopy (XAS). Molybdenum K-edge XAS data were acquired from several inactive simulant high level nuclear waste

  16. Molybdenum in Nuclear Waste Glasses -Incorporation and Redox state R.J. Short, R.J. Hand, N.C. Hyatt,

    E-Print Network [OSTI]

    Sheffield, University of

    Molybdenum in Nuclear Waste Glasses - Incorporation and Redox state R.J. Short, R.J. Hand, N form in simulated high level nuclear waste (HLW) glass melts have been studied. It was found less attention has been paid to the effects of redox on nuclear waste glasses. One particular element

  17. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

    1999-10-01T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  18. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    SciTech Connect (OSTI)

    A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

    1999-09-30T23:59:59.000Z

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  19. Scoping evaluation of the technical capabilities of DOE sites for disposal of hazardous metals in mixed low-level waste

    SciTech Connect (OSTI)

    Gruebel, M.M.; Waters, R.D.; Langkopf, B.S.

    1997-05-01T23:59:59.000Z

    A team of analysts designed and conducted a scoping evaluation to estimate the technical capabilities of fifteen Department of Energy sites for disposal of the hazardous metals in mixed low-level waste (i.e., waste that contains both low-level radioactive materials and hazardous constituents). Eight hazardous metals were evaluated: arsenic, barium, cadmium, chromium, lead, mercury, selenium, and silver. The analysis considered transport only through the groundwater pathway. The results are reported as site-specific estimates of maximum concentrations of each hazardous metal in treated mixed low-level waste that do not exceed the performance measures established for the analysis. Also reported are site-specific estimates of travel times of each hazardous metal to the point of compliance.

  20. Next Generation Nuclear Plant Methods Technical Program Plan

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2007-01-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  1. Next Generation Nuclear Plant Methods Technical Program Plan

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2010-12-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  2. Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2010-09-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  3. CORROSION OF NUCLEAR WASTE GLASSES IN NON-SATURATED CONDITIONS: TIME-TEMPERATURE BEHAVIOUR

    E-Print Network [OSTI]

    Sheffield, University of

    CORROSION OF NUCLEAR WASTE GLASSES IN NON-SATURATED CONDITIONS: TIME-TEMPERATURE BEHAVIOUR Michael Rostovsky Lane, 2/14, Moscow, 119121, Russia This paper examines corrosion behaviour of radioactive term natural tests. These demonstrated diminishing of release rates of radionuclides by time. Corrosion

  4. Radionuclide transport code development in support of nuclear waste storage investigations

    SciTech Connect (OSTI)

    Martinez, M.J.; Bixler, N.E.

    1983-03-01T23:59:59.000Z

    This report documents the status of radionuclide transport code development in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) program as of October 1982. The modifications made to an existing code, FEMWASTE, are described and subsequent verification of the modified code is presented.

  5. An underground characterization program for a nuclear fuel waste disposal vault in plutonic rock

    SciTech Connect (OSTI)

    Thompson, P.M.; Everitt, R.A. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1993-12-31T23:59:59.000Z

    The Canadian Nuclear Fuel Waste Management Program (CNFWMP) is developing a concept for disposing of nuclear fuel waste that involves placing and sealing it in a disposal vault excavated 500 to 1,000 m deep in the stable plutonic rock of the Canadian Shield. In this concept, engineered and natural barriers serve to isolate the waste from the biosphere. Since 1983, underground characterization and testing in support of the CNFWMP has been ongoing at the Underground Research Laboratory (URL) in southeastern Manitoba. This paper draws on experience gained at the URL to recommend an approach to underground characterization that would provide the necessary information to make design decisions for a disposal vault in plutonic rock.

  6. TECHNICAL ASSESSMENT OF BULK VITRIFICATION PROCESS & PRODUCT FOR TANK WASTE TREATMENT AT THE DEPARTMENT OF ENERGY HANFORD SITE

    SciTech Connect (OSTI)

    SCHAUS, P.S.

    2006-07-21T23:59:59.000Z

    At the U.S. Department of Energy (DOE) Hanford Site, the Waste Treatment Plant (WTP) is being constructed to immobilize both high-level waste (IUW) for disposal in a national repository and low-activity waste (LAW) for onsite, near-surface disposal. The schedule-controlling step for the WTP Project is vitrification of the large volume of LAW, current capacity of the WTP (as planned) would require 50 years to treat the Hanford tank waste, if the entire LAW volume were to be processed through the WTP. To reduce the time and cost for treatment of Hanford Tank Waste, and as required by the Tank Waste Remediation System Environmental Impact Statement Record of Decision and the Hanford Federal Facility Consent Agreement (Tn-Party Agreement), DOE plans to supplement the LAW treatment capacity of the WTP. Since 2002, DOE, in cooperation with the Environmental Protection Agency and State of Washington Department of Ecology has been evaluating technologies that could provide safe and effective supplemental treatment of LAW. Current efforts at Hanford are intended to provide additional information to aid a joint agency decision on which technology will be used to supplement the WTP. A Research, Development and Demonstration permit has been issued by the State of Washington to build and (for a limited time) operate a Demonstration Bulk Vitrification System (DBVS) facility to provide information for the decision on a supplemental treatment technology for up to 50% of the LAW. In the Bulk Vitrification (BV) process, LAW, soil, and glass-forming chemicals are mixed, dried, and placed in a refractory-lined box, Electric current, supplied through two graphite electrodes in the box, melts the waste feed, producing a durable glass waste-form. Although recent modifications to the process have resulted in significant improvements, there are continuing technical concerns.

  7. Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.

    SciTech Connect (OSTI)

    Liu, Haiqing; Bell, Nelson Simmons; Cipiti, Benjamin B.; Lewis, Tom Goslee,; Sava, Dorina Florentina; Nenoff, Tina Maria

    2012-09-01T23:59:59.000Z

    Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

  8. Solid waste integrated forecast technical (SWIFT) report: FY1997 to FY 2070, Revision 1

    SciTech Connect (OSTI)

    Valero, O.J.; Templeton, K.J.; Morgan, J.

    1997-01-07T23:59:59.000Z

    This web site provides an up-to-date report on the radioactive solid waste expected to be managed by Hanford's Waste Management (WM) Project from onsite and offsite generators. It includes: an overview of Hanford-wide solid waste to be managed by the WM Project; program-level and waste class-specific estimates; background information on waste sources; and comparisons with previous forecasts and with other national data sources. This web site does not include: liquid waste (current or future generation); waste to be managed by the Environmental Restoration (EM-40) contractor (i.e., waste that will be disposed of at the Environmental Restoration Disposal Facility (ERDF)); or waste that has been received by the WM Project to date (i.e., inventory waste). The focus of this web site is on low-level mixed waste (LLMW), and transuranic waste (both non-mixed and mixed) (TRU(M)). Some details on low-level waste and hazardous waste are also provided. Currently, this web site is reporting data th at was requested on 10/14/96 and submitted on 10/25/96. The data represent a life cycle forecast covering all reported activities from FY97 through the end of each program's life cycle. Therefore, these data represent revisions from the previous FY97.0 Data Version, due primarily to revised estimates from PNNL. There is some useful information about the structure of this report in the SWIFT Report Web Site Overview.

  9. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    SciTech Connect (OSTI)

    Not Available

    1990-03-01T23:59:59.000Z

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs.

  10. A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

    SciTech Connect (OSTI)

    Carelli, M.D.; Franceschini, F.; Lahoda, E.J. [Westinghouse Electric Company LLC., Cranberry Township, PA (United States); Petrovic, B. [Georgia Institute of Technology, Atlanta, GA (United States)

    2012-07-01T23:59:59.000Z

    A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

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

    SciTech Connect (OSTI)

    M. D. Staiger

    1999-06-01T23:59:59.000Z

    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.

  12. Guideline for the seismic technical evaluation of replacement items for nuclear power plants

    SciTech Connect (OSTI)

    Harris, S.P.; Cushing, R.W. (EQE International, San Francisco, CA (United States)); Johnson, H.W. (Programmatic Solutions, Smithtown, NY (United States)); Abeles, J.M. (System 1, Inc., Potomac, MD (United States))

    1993-02-01T23:59:59.000Z

    Seismic qualification for equipment originally installed in nuclear power plants was typically performed by the original equipment suppliers or manufactures (OES/OEM). Many of the OES/OEM no longer maintain quality assurance programs with adequate controls for supplying nuclear equipment. Utilities themselves must provide reasonable assurance in the continued seismic adequacy of such replacement items. This guideline provides practical, cost-effective techniques which can be used to provide reasonable assurance that replacement items will meet seismic performance requirements necessary to maintain the seismic design basis of commercial nuclear power plants. It also provides a method for determining when a seismic technical evaluation of replacement items (STERI) is required as part of the procurement process for spare and replacement items. Guidance on supplier program requirements necessary to maintain continued seismic adequacy and on documentation of maintaining required seismic adequacy is also included.

  13. RADIOACTIVE WASTE MANAGEMENT IN THE CHERNOBYL EXCLUSION ZONE - 25 YEARS SINCE THE CHERNOBYL NUCLEAR POWER PLANT ACCIDENT

    SciTech Connect (OSTI)

    Farfan, E.; Jannik, T.

    2011-10-01T23:59:59.000Z

    Radioactive waste management is an important component of the Chernobyl Nuclear Power Plant accident mitigation and remediation activities of the so-called Chernobyl Exclusion Zone. This article describes the localization and characteristics of the radioactive waste present in the Chernobyl Exclusion Zone and summarizes the pathways and strategy for handling the radioactive waste related problems in Ukraine and the Chernobyl Exclusion Zone, and in particular, the pathways and strategies stipulated by the National Radioactive Waste Management Program. The brief overview of the radioactive waste issues in the ChEZ presented in this article demonstrates that management of radioactive waste resulting from a beyond-designbasis accident at a nuclear power plant becomes the most challenging and the costliest effort during the mitigation and remediation activities. The costs of these activities are so high that the provision of radioactive waste final disposal facilities compliant with existing radiation safety requirements becomes an intolerable burden for the current generation of a single country, Ukraine. The nuclear accident at the Fukushima-1 NPP strongly indicates that accidents at nuclear sites may occur in any, even in a most technologically advanced country, and the Chernobyl experience shows that the scope of the radioactive waste management activities associated with the mitigation of such accidents may exceed the capabilities of a single country. Development of a special international program for broad international cooperation in accident related radioactive waste management activities is required to handle these issues. It would also be reasonable to consider establishment of a dedicated international fund for mitigation of accidents at nuclear sites, specifically, for handling radioactive waste problems in the ChEZ. The experience of handling Chernobyl radioactive waste management issues, including large volumes of radioactive soils and complex structures of fuel containing materials can be fairly useful for the entire world's nuclear community and can help make nuclear energy safer.

  14. Technical papers presented at the Defense Nuclear Agency Global Effects Review. Held at Moffett Field, California on 25-27 February 1986. Volume 3. Technical report, 25 February-12 May 1986

    SciTech Connect (OSTI)

    Not Available

    1986-05-15T23:59:59.000Z

    This document contains technical papers presented at the Defense Nuclear Agency Review of Global Effects held at NASA Ames Research Center 25-027 February 1986.

  15. Technical papers presented at the Defense Nuclear Agency Global Effects Review. Held at Moffett Field, California on 25-29 February 1986. Volume 2. Technical report, 25 February-12 May 1986

    SciTech Connect (OSTI)

    Not Available

    1986-05-15T23:59:59.000Z

    This document contains technical papers presented at the defense Nuclear Agency Review of Global Effects held at NASA Ames Research Center 25-27 February 1986.

  16. Development of Vitrification Process and Glass Formulation for Nuclear Waste Conditioning

    SciTech Connect (OSTI)

    Petitjean, V.; Fillet, C.; Boen, R.; Veyer, C.; Flament, T.

    2002-02-26T23:59:59.000Z

    The vitrification of high-level waste is the internationally recognized standard to minimize the impact to the environment resulting from waste disposal as well as to minimize the volume of conditioned waste to be disposed of. COGEMA has been vitrifying high-level waste industrially for over 20 years and is currently operating three commercial vitrification facilities based on a hot metal crucible technology, with outstanding records of safety, reliability and product quality. To further increase the performance of vitrification facilities, CEA and COGEMA have been developing the cold crucible melter technology since the beginning of the 1980s. This type of melter is characterized by a virtually unlimited equipment service life and a great flexibility in dealing with various types of waste and allowing development of high temperature matrices. In complement of and in parallel with the vitrification process, a glass formulation methodology has been developed by the CEA in order to tailor matrices for the wastes to be conditioned while providing the best adaptation to the processing technology. The development of a glass formulation is a trade-off between material properties and qualities, technical feasibility, and disposal safety criteria. It involves non-radioactive and radioactive laboratories in order to achieve a comprehensive matrix qualification. Several glasses and glass ceramics have thus been studied by the CEA to be compliant with industrial needs and waste characteristics: glasses or other matrices for a large spectrum of fission products, or for high contents of specifics elements such as sodium, phosphate, iron, molybdenum, or actinides. New glasses or glass-ceramics designed to minimize the final wasteform volume for solutions produced during the reprocessing of high burnup fuels or to treat legacy wastes are now under development and take benefit from the latest CEA hot-laboratories and technology development. The paper presents the CEA state-of-the-art in developing matrices or glasses and provides several examples.

  17. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    Cnalk R i v e r , Ontar o, AECL-b308, 46 p . Tnomas, R. K. ,with a radioactive waste vault. AECL--6308, Atomic Energys t u d y . P a r t 2. AECL-6188-2, Atomic Energy of Canada,

  18. Site characterization progress report: Yucca Mountain, Nevada, October 1, 1994--March 31, 1995, Number 12. Nuclear Waste Policy Act (Section 113)

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    During the first half of fiscal year 1995, most activities at the Yucca Mountain Site Characterization Project were directed at implementing the Program Plan developed by the Office of Civilian Radioactive Waste Management. The Plan is designed to enable the Office to make measurable and significant progress toward key objectives over the next five years within the financial resources that can be realistically expected. Activities this period focused on the immediate goal of determining by 1998 whether Yucca Mountain, Nevada, is technically suitable as a possible site for a geologic repository for the permanent disposal of spent nuclear fuel and high-level radioactive waste. Work on the Project advanced in several critical areas, including programmatic activities such as issuing the Program Plan, completing the first technical basis report to support the assessment of three 10 CFR 960 guidelines, developing the Notice of Intent for the Environmental Impact Statement, submitting the License Application Annotated Outline, and beginning a rebaselining effort to conform with the goals of the Program Plan. Scientific investigation and analysis of the site and design and construction activities to support the evaluation of the technical suitability of the site also advanced. Specific details relating to all Project activities and reports generated are presented in this report.

  19. Nye County, Nevada 1992 nuclear waste repository program: Program overview. Final report

    SciTech Connect (OSTI)

    NONE

    1998-07-01T23:59:59.000Z

    The purpose of this document is to provide an overview of the Nye County FY92 Nuclear Waste Repository Program (Program). Funds to pay for Program costs will come from the Federal Nuclear Waste Fund, which was established under the Nuclear Waste Policy Act of 1982 (NWPA). In early 1983, the Yucca Mountain was identified as a potentially suitable site for the nation`s first geologic repository for spent reactor fuel and high-level radioactive waste. Later that year, the Nye County Board of County Commissioners (Board) established the capability to monitor the Federal effort to implement the NWPA and evaluate the potential impacts of repository-related activities on Nye County. Over the last eight years, the County`s program has grown in complexity and cost in order to address DOE`s evolving site characterization studies, and prepare for the potential for facility construction and operation. Changes were necessary as well, in response to Congress`s redirection of the repository program specified in the amendments, to the NWPA approved in 1987. In early FY 1991, the County formally established a project office to plan and implement its program of work. The Repository Project Office`s (RPO) mission and functions are provided in Section 2.0. The RPO organization structure is described in Section 3.0.

  20. A review of the environmental survivability of telerobotic control sensor systems for use in nuclear waste tanks

    SciTech Connect (OSTI)

    Holcomb, D.E.; Burks, B.L.

    1994-05-01T23:59:59.000Z

    This report was prepared by the Oak Ridge National Laboratory (ORNL) and funded by the Department of Energy (DOE) Office of Technology Development (OTD) Robotics Technology Development Program (RTDP). During the next few years field deployment of remotely operated systems in nuclear waste cleanup operations will increase dramatically as DOE strives to efficiently and safely remediate the many waste storage sites. Typically, the most fragile components in remote systems are the sensors that provide feedback to the operators or to computer control algorithms. The purpose of this review is to determine the availability of environmentally hardened sensors to support control of a manipulator or vehicle system in a waste tank environment. The emphasis of the report is on the environmental ruggedness of currently available sensors. For the purpose of this review a set of nominal requirements for survivability were adopted conditions in the single-shell tanks at Hanford. This report is designed to be a practical guide to the state of the art in commercially available environmentally tolerant sensors for use with robotic systems. It is neither intended to be an exhaustive review of the technical literature on potential measurement techniques nor a complete physical review of the functioning of particular sensor systems. This report is intended to be a living document. As additional, corrected, or updated information is received from sensor manufacturers, it will be incorporated into the report database. The physical report will then be periodically revised and released in updated format. The authors wish to apologize to any sources of environmentally hardened sensors that were omitted during this review and encourage submission of new or updated data.

  1. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

    1984-01-01T23:59:59.000Z

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

  2. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

    1982-03-31T23:59:59.000Z

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

  3. Volume reduction and vitrification of nuclear waste with thermal plasma

    SciTech Connect (OSTI)

    Hoffelner, W. [Moser-Glaser and Co., Muttenz (Switzerland); Chrubasik, A. [NUKEM GmbH, Alzenau (Germany); Eschenbach, R.C. [RETECH Inc., Ukiah, CA (United States)

    1993-12-31T23:59:59.000Z

    A process for efficient and safe destruction of organics and vitrification of low/medium level radioactive waste is presented. A transferred arc plasma torch is employed as the heat source. The process handles several types of feed: combustibles, inorganic materials and metals. A non-leaching glassy solid which can be stored without further treatment is obtained as the final product. High volume-reduction factors can be achieved with this process. A wet gas cleaning system leads to extremely clean off-gas.

  4. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    NONE

    1996-12-01T23:59:59.000Z

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens.

  5. An international initiative on long-term behavior of high-level nuclear waste glass

    SciTech Connect (OSTI)

    Gin, Stephane [CEA Marcoule DTCD SECM LCLT, Bagnols/Ceze (France); Abdelouas, Abdessalam [SUBATECH, Nantes (France); Criscenti, Louise J. [Sandia National Laboratories, Albuquerque, NM (United States); Ebert, W. L. [Argonne National Laboratory (ANL), Argonne, IL (United States); Ferrand, Karine [SCK·CEN, Mol (Belgium); Geisler, Thorsten [Rheinische Friedrich-Wilhelms-Univ., Bonn (Germany); Harrison, Mike T. [National Nuclear Laboratory, Sellafield, Cumbria (United Kingdom); Inagaki, Yaohiro [Kyushu Univ. (Japan). Dept. Appl. Quantum Physics and Nuclear Engineering; Mitsui, Seiichiro [Japan Atomic Energy Agency, Ibaraki (Japan); Mueller, Karl T. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States). Environmental and Molecular Science Lab.; Marra, James C. [Savannah River National Laboratory, Aiken, SC (United States); Pantano, Carlo G. [Penn State Univ., State College, PA (United States); Pierce, Eric M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Ryan, Joseph V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Schofield, James M. [AMEC, Harwell Oxford (United Kingdom); Steefel, Carl I. [Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States). Earth Sciences Div.; Vienna, John D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2013-06-01T23:59:59.000Z

    Nations producing borosilicate glass as an immobilization material for radioactive wastes resulting from spent nuclear fuel reprocessing have reinforced scientific collaboration to obtain consensus on mechanisms controlling the long-term dissolution rate of glass. This goal is deemed to be crucial for the development of reliable performance assessment models for geological disposal. The collaborating laboratories all conduct fundamental and/or applied research with modern materials science techniques. The paper briefly reviews the radioactive waste vitrification programmes of the six participant nations and summarizes the state-of-the-art of glass corrosion science, emphasizing common scientific needs and justifications for on-going initiatives.

  6. Extended Dry Storage of Used Nuclear Fuel: Technical Issues: A USA Perspective

    SciTech Connect (OSTI)

    McConnell, Paul; Hanson, Brady D.; Lee, Moo; Sorenson, Ken B.

    2011-10-28T23:59:59.000Z

    Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-to-safety structures, systems and components (SSC's) continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSC's. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

  7. Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems

    SciTech Connect (OSTI)

    Doolen, G.D.; Venneri, F.; Li, N.; Williamson, M.A.; Houts, M.; Lawrence, G.

    1998-06-27T23:59:59.000Z

    ATW destroys virtually all the plutonium and higher actinides without reprocessing the spent fuel in a way that could lead to weapons material diversion. An ATW facility consists of three major elements: (1) a high-power proton linear accelerator; (2) a pyrochemical spent fuel treatment i waste cleanup system; (3) a liquid lead-bismuth cooled burner that produces and utilizes an intense source-driven neutron flux for transmutation in a heterogeneous (solid fuel) core. The concept is the result of many years of development at LANL as well as other major international research centers. Once demonstrated and developed, ATW could be an essential part of a global non-proliferation strategy for countries that could build up large quantities of plutonium from their commercial reactor waste. ATW technology, initially proposed in the US, has received wide and rapidly increasing attention abroad, especially in Europe and the Far East with major programs now being planned, organized and tided. Substantial convergence presently exists on the technology choices among the programs, opening the possibility of a strong and effective international collaboration on the phased development of the ATW technology.

  8. Truck and rail charges for shipping spent fuel and nuclear waste

    SciTech Connect (OSTI)

    McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

    1986-06-01T23:59:59.000Z

    The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials.

  9. I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

    SciTech Connect (OSTI)

    S. Frank

    2010-09-01T23:59:59.000Z

    The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of in the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests were produced concentrated salt products that acted as the feed material for development of advanced waste forms investigated in this proposal. Accomplishments from the first year activities associated with this I-NERI project included the down selection of candidate waste forms to immobilize fission products separated from electrorefiner salt, and the design of equipment to fabricate actual waste forms in the Hot Fuels Examination Facility (HFEF) at the INL. Reported in this document are accomplishments from the second year (FY10) work performed at the INL, and includes the testing of waste form fabrication equipment, repeating the fission product precipitation experiment, and initial waste form fabrication efforts.

  10. Decontamination of Nuclear Liquid Wastes Status of CEA and AREVA R and D: Application to Fukushima Waste Waters - 12312

    SciTech Connect (OSTI)

    Fournel, B.; Barre, Y.; Lepeytre, C.; Peycelon, H. [CEA Marcoule, DTCD, BP17171, 30207 Bagnols sur Ceze (France); Grandjean, A. [Institut de Chimie Separative de Marcoule, UMR5257 CEA-CNRS-UM2-ENSCM, BP17171, 30207 Bagnols sur Ceze (France); Prevost, T.; Valery, J.F. [AREVA NC, Paris La Defense (France); Shilova, E.; Viel, P. [CEA Saclay, DSM/IRAMIS/SPCSI, 91191 Gif sur Yvette (France)

    2012-07-01T23:59:59.000Z

    Liquid wastes decontamination processes are mainly based on two techniques: Bulk processes and the so called Cartridges processes. The first technique has been developed for the French nuclear fuel reprocessing industry since the 60's in Marcoule and La Hague. It is a proven and mature technology which has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The second technique, involving cartridges processes, offers new opportunities for the use of innovative adsorbents. The AREVA process developed for Fukushima and some results obtained on site will be presented as well as laboratory scale results obtained in CEA laboratories. Examples of new adsorbents development for liquid wastes decontamination are also given. A chemical process unit based on co-precipitation technique has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The asset of this technique is its ability to process large volumes in a continuous mode. Several chemical products can be used to address specific radioelements such as: Cs, Sr, Ru. Its drawback is the production of sludge (about 1% in volume of initial liquid volume). CEA developed strategies to model the co-precipitation phenomena in order to firstly minimize the quantity of added chemical reactants and secondly, minimize the size of co-precipitation units. We are on the way to design compact units that could be mobilized very quickly and efficiently in case of an accidental situation. Addressing the problem of sludge conditioning, cementation appears to be a very attractive solution. Fukushima accident has focused attention on optimizations that should be taken into account in future studies: - To better take account for non-typical aqueous matrixes like seawater; - To enlarge the spectrum of radioelements that can be efficiently processed and especially short lives radioelements that are usually less present in standard effluents resulting from nuclear activities; - To develop reversible solid adsorbents for cartridge-type applications in order to minimize wastes. (authors)

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

    SciTech Connect (OSTI)

    Staiger, M. Daniel, Swenson, Michael C.

    2011-09-01T23:59:59.000Z

    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.

  12. Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel

    SciTech Connect (OSTI)

    Crum, Jarrod V.; Strachan, Denis M.; Rohatgi, Aashish; Zumhoff, Mac R.

    2013-10-01T23:59:59.000Z

    Epsilon metal (?-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000°C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

  13. Milestones for selection, characterization, and analysis of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain.

    SciTech Connect (OSTI)

    Rechard, Robert P.

    2014-02-01T23:59:59.000Z

    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.

  14. Spent Nuclear Fuel Project technical baseline document. Fiscal year 1995: Volume 1, Baseline description

    SciTech Connect (OSTI)

    Womack, J.C. [Westinghouse Hanford Co., Richland, WA (United States); Cramond, R. [TRW (United States); Paedon, R.J. [SAIC (United States)] [and others

    1995-03-13T23:59:59.000Z

    This document is a revision to WHC-SD-SNF-SD-002, and is issued to support the individual projects that make up the Spent Nuclear Fuel Project in the lower-tier functions, requirements, interfaces, and technical baseline items. It presents results of engineering analyses since Sept. 1994. The mission of the SNFP on the Hanford site is to provide safety, economic, environmentally sound management of Hanford SNF in a manner that stages it to final disposition. This particularly involves K Basin fuel, although other SNF is involved also.

  15. Nuclear-waste encapsulation by metal-matrix casting

    SciTech Connect (OSTI)

    Nelson, R.G.; Nesbitt, J.F.; Slate, S.C.

    1981-05-01T23:59:59.000Z

    Several encapsulation casting processes are described that were developed or used at the Pacific Northwest Laboratory to embed simulated high-level wastes of two different forms (glass marbles and ceramic pellets) in metal matrices. Preliminary evaluations of these casting processes and the products are presented. Demonstrations have shown that 5- to 10-mm-dia glass marbles can be encapsulated on an engineering scale with lead or lead alloys by gravity or vacuum processes. Marbles approx. 12 mm in dia were successfully encapsulated in a lead alloy on a production scale. Also, 4- to 9-mm-dia ceramic pellets in containers of various sizes were completely penetrated and the individual pellets encased with aluminum-12 wt % silicon alloy by vacuum processes. Indications are that of the casting processes tested, aluminum 12 wt % silicon alloy vacuum-cast around ceramic pellets had the highest degree of infiltration or coverage of pellet surfaces.

  16. Heat pipe effects in nuclear waste isolation: a review

    SciTech Connect (OSTI)

    Doughty, C.; Pruess, K.

    1985-12-01T23:59:59.000Z

    The existence of fractures favors heat pipe development in a geologic repository as does a partially saturated medium. A number of geologic media are being considered as potential repository sites. Tuff is partially saturated and fractured, basalt and granite are saturated and fractured, salt is unfractured and saturated. Thus the most likely conditions for heat pipe formation occur in tuff while the least likely occur in salt. The relative permeability and capillary pressure dependences on saturation are of critical importance for predicting thermohydraulic behavior around a repository. Mineral redistribution in heat pipe systems near high-level waste packages emplaced in partially saturated formations may significantly affect fluid flow and heat transfer processes, and the chemical environment of the packages. We believe that a combined laboratory, field, and theoretical effort will be needed to identify the relevant physical and chemical processes, and the specific parameters applicable to a particular site. 25 refs., 1 fig.

  17. ANL Technical Support Program for DOE Environmental Restoration and Waste Management; Annual report, October 1992--September 1993

    SciTech Connect (OSTI)

    Bates, J.K.; Bourcier, W.L.; Bradley, C.R. [and others

    1994-06-01T23:59:59.000Z

    This report is an overview of the progress during FY 1993 for the Technical Support Program that is part of the ANL Technology Support Activity for DOE Environmental Restoration and Waste Management (EM). The purpose is to evaluate, before hot start-up of the Defense Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP), factors that are anticipated to affect glass reaction in an unsaturated environment typical of what may be expected for the candidate Yucca Mountain repository site. Specific goals for the testing program include the following: reviewing and evaluating available data on parameters that will be important in establishing the long-term performance of glass in a repository environment; performing tests to further quantify the effects of important variables where there are deficiencies in the available data; and initiating long-term tests to determine glass performance under a range of conditions applicable to repository disposal.

  18. Assessing the Feasibility of Interrogating Nuclear Waste Storage Silos using Cosmic-ray Muons

    E-Print Network [OSTI]

    Ambrosino, F; Cimmino, L; D'Alessandro, R; Ireland, D G; Kaiser, R; Mahon, D F; Mori, N; Noli, P; Saracino, G; Shearer, C; Viliani, L; Yang, G

    2014-01-01T23:59:59.000Z

    Muon radiography is a fast growing field in applied scientific research. In recent years, many detector technologies and imaging techniques using the Coulomb scattering and absorption properties of cosmic-ray muons have been developed for the non-destructive assay of various structures across a wide range of applications. This work presents the first results that assess the feasibility of using muons to interrogate waste silos within the UK Nuclear Industry. Two such approaches, using different techniques that exploit each of these properties, have previously been published, and show promising results from both simulation and experimental data for the detection of shielded high-Z materials and density variations from volcanic assay. Both detector systems are based on scintillator and photomultiplier technologies. Results from dedicated simulation studies using both these technologies and image reconstruction techniques are presented for an intermediate-sized nuclear waste storage facility filled with concrete...

  19. Politics and promises of nuclear waste disposal: the view from Nevada

    SciTech Connect (OSTI)

    Bryan, R.H.

    1987-10-01T23:59:59.000Z

    DOE`s betrayal of the principles and standards of the Nuclear Waste Policy Act (NWPA) has distorted the agency`s repository-siting decisions. Leadership is needed to make midcourse corrections and to return to the promise of state-federal cooperation on which the act was built. NWPA managed to incorporate the interests of diverse factions into a decision-making process that was viewed as an equitable and workable solution to the nation`s nuclear waste disposal dilemma. The House of Representatives subcommittees report documents conclusively a substantial and pervasive bias in favor of the selection of sites at Yucca Mountain and Hanford and a politization of the siting process.

  20. What are Spent Nuclear Fuel and High-Level Radioactive Waste ?

    SciTech Connect (OSTI)

    DOE

    2002-12-01T23:59:59.000Z

    Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

  1. The Potential of Cellulosic Ethanol Production from Municipal Solid Waste: A Technical and Economic Evaluation

    E-Print Network [OSTI]

    Shi, Jian; Ebrik, Mirvat; Yang, Bin; Wyman, Charles E.

    2009-01-01T23:59:59.000Z

    process streams. Handb. Bioethanol:395-415. 10. Ehrman T.solid waste used as bioethanol sources and its relatedof cellulosic biomass into bioethanol as an alternative

  2. Application of Microprocessor-Based Equipment in Nuclear Power Plants - Technical Basis for a Qualification Methodology

    SciTech Connect (OSTI)

    Korsah, K.

    2001-08-24T23:59:59.000Z

    This document (1) summarizes the most significant findings of the ''Qualification of Advanced Instrumentation and Control (I&C) Systems'' program initiated by the Nuclear Regulatory Commission (NRC); (2) documents a comparative analysis of U.S. and European qualification standards; and (3) provides recommendations for enhancing regulatory guidance for environmental qualification of microprocessor-based safety-related systems. Safety-related I&C system upgrades of present-day nuclear power plants, as well as I&C systems of Advanced Light-Water Reactors (ALWRs), are expected to make increasing use of microprocessor-based technology. The Nuclear Regulatory Commission (NRC) recognized that the use of such technology may pose environmental qualification challenges different from current, analog-based I&C systems. Hence, it initiated the ''Qualification of Advanced Instrumentation and Control Systems'' program. The objectives of this confirmatory research project are to (1) identify any unique environmental-stress-related failure modes posed by digital technologies and their potential impact on the safety systems and (2) develop the technical basis for regulatory guidance using these findings. Previous findings from this study have been documented in several technical reports. This final report in the series documents a comparative analysis of two environmental qualification standards--Institute of Electrical and Electronics Engineers (IEEE) Std 323-1983 and International Electrotechnical Commission (IEC) 60780 (1998)--and provides recommendations for environmental qualification of microprocessor-based systems based on this analysis as well as on the findings documented in the previous reports. The two standards were chosen for this analysis because IEEE 323 is the standard used in the U.S. for the qualification of safety-related equipment in nuclear power plants, and IEC 60780 is its European counterpart. In addition, the IEC document was published in 1998, and should reflect any new qualification concerns, from the European perspective, with regard to the use of microprocessor-based safety systems in power plants.

  3. Sub-Seabed Repository for Nuclear Waste - a Strategic Alternative - 13102

    SciTech Connect (OSTI)

    McAllister, Keith R. [Department of the Navy, 15 Turkey Foot Court, Darnestown, MD 20878 (United States)] [Department of the Navy, 15 Turkey Foot Court, Darnestown, MD 20878 (United States)

    2013-07-01T23:59:59.000Z

    It was recognized at the outset of nuclear power generation in the 1950's that the waste products would require isolation away from humans for periods in excess of 10,000 years. After years studying alternatives, the DOE recommended pursuing the development of a SNF/HLW disposal facility within Yucca Mountain in the desert of Nevada. That recommendation became law with passage of the NWPAA, effectively stopping development of other approaches to the waste problem. In the face of political resistance from the state of Nevada, the 2010 decision to withdraw the license application for the geologic repository at Yucca Mountain has delayed further the most mature option for safe, long-term disposal of SNF and HLW. It is time to revisit an alternative option, sub-seabed disposal within the US Exclusive Economic Zone (EEZ), which would permanently sequester waste out of the biosphere, and out of the reach of saboteurs or terrorists. A proposal is made for a full scale pilot project to demonstrate burying radioactive waste in stable, deep ocean sediments. While much of the scientific work on pelagic clays has been done to develop a sub-seabed waste sequestration capability, this proposal introduces technology from non-traditional sources such as riser-less ocean drilling and the Navy's Sound Surveillance System. The political decisions affecting the issue will come down to site selection and a thorough understanding of comparative risks. The sub-seabed sequestration of nuclear waste has the potential to provide a robust solution to a critical problem for this clean and reliable energy source. (authors)

  4. The NOx system in nuclear waste. 1997 annual progress report

    SciTech Connect (OSTI)

    Meisel, D. [Argonne National Lab., IL (US). Chemistry Div.; Camaioni, D.; Orlando, T. [Pacific Northwest National Lab., Richland, WA (US)

    1997-01-01T23:59:59.000Z

    'The authors highlight their results from the title project. The project is a coordinated effort of the three Co-PIs to assist the Safety Programs at the Hanford and other DOE Environmental Management Sites. The authors present in the report their observations and interactively discuss their implications for safety concerns. They focus on three issues: (1) Reducing radicals in the NOx system The authors show that the only reducing radical that lasts longer than a few ns in typical waste solutions, and is capable of generating hydrogen, is NO{sub 3}{sup 2-}. The authors measured the lifetime of this species across the whole pH range (3 {le} pH {le} 14) and found it to be shorter than -15 \\265s, before it dissociates to give the strongly oxidizing NO, radicals. They found that it reacts with many proton donors (H{sup +}, phosphate, borate, NH{prime}, amines) in a reaction that is not merely an acid-base equilibrium reaction but is probably a dissociative proton transfer. They estimate the redox potential from theoretical considerations and obtain an experimental verification. They conclude that it is highly unlikely, although thermodynamically possible, that this radi-cal will generate hydrogen in waste solutions. (2) Aging of organic chelators and their degradation products by NO, Methodologies to study the degradation of organic substrates (including the important waste components, formate and oxalate) to CO;, or carbonate, by NO, were developed. This radical dimerizes and disproportionates to nitrate and nitrite. Therefore, mineraliza-tion of the organic substrates competes with the disproportionation of NO,. Among the organic substrates, formate and oxalate are also mineralized but because they are of low fuel value their mineralization is not very helpful, yet it consumes NO,. (3) Interfacial processes in aqueous suspensions Yields of charge transfer from solid silica particles to water and other liquids were meas-ured. If the particles are small enough, essentially all of the charge that is originally depos-ited in the solid escapes into the liquid. This implies that the solid/liquid interface does not provide a significant barrier to the transfer of charges into the solution when the particles are very small (I 20 nm). Electrons may reach the liquid and generate hydrogen, for example. On the other hand, the same mechanism may also provide a pathway for oxidative aging of organics by holes even when the organic is dissolved in the liquid or adsorbed on the solid surface. The authors have started to study reactions of NO,. Methodology and instrumentation to measure reactions of relevant organic radicals with NO, and with its parent NO, were developed. Because of low extinction coefficients, conductivity will be the method of choice.'

  5. The strengthening and repair of underground structures: A new approach to the management of nuclear waste

    SciTech Connect (OSTI)

    Colgate, S.A.

    1991-01-01T23:59:59.000Z

    This paper presents three closely related ideas and technologies: (1) The secure, repairable, long time confinement of nuclear radioactive waste underground by a large surrounding region of compressive overstress; (2) The inherent tectonic weakness and vulnerability of the normal underground environment and its modification by overstress; (3) The process of creating overstress by the sequential periodic high pressure injection of a finite gel strength rapid setting grout. 12 refs., 6 figs.

  6. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    f o r 1 0 - y e a r - o l d PWR SF. I n i t i a l Heat G e nt nuclear fuel cycles for a PWR. Decay heat power for d i fKBS LWR MOX NRC NWTS ONWI OWI PWR RH-TRU WIPP b o i l i n g

  7. West Valley high-level nuclear waste glass development: a statistically designed mixture study

    SciTech Connect (OSTI)

    Chick, L.A.; Bowen, W.M.; Lokken, R.O.; Wald, J.W.; Bunnell, L.R.; Strachan, D.M.

    1984-10-01T23:59:59.000Z

    The first full-scale conversion of high-level commercial nuclear wastes to glass in the United States will be conducted at West Valley, New York, by West Valley Nuclear Services Company, Inc. (WVNS), for the US Department of Energy. Pacific Northwest Laboratory (PNL) is supporting WVNS in the design of the glass-making process and the chemical formulation of the glass. This report describes the statistically designed study performed by PNL to develop the glass composition recommended for use at West Valley. The recommended glass contains 28 wt% waste, as limited by process requirements. The waste loading and the silica content (45 wt%) are similar to those in previously developed waste glasses; however, the new formulation contains more calcium and less boron. A series of tests verified that the increased calcium results in improved chemical durability and does not adversely affect the other modeled properties. The optimization study assessed the effects of seven oxide components on glass properties. Over 100 melts combining the seven components into a wide variety of statistically chosen compositions were tested. Viscosity, electrical conductivity, thermal expansion, crystallinity, and chemical durability were measured and empirically modeled as a function of the glass composition. The mathematical models were then used to predict the optimum formulation. This glass was tested and adjusted to arrive at the final composition recommended for use at West Valley. 56 references, 49 figures, 18 tables.

  8. Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532

    SciTech Connect (OSTI)

    Gasbarro, Christina; Bello, Job [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States)] [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States); Bryan, Samuel; Lines, Amanda; Levitskaia, Tatiana [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)] [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)

    2013-07-01T23:59:59.000Z

    Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)

  9. Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste

    SciTech Connect (OSTI)

    Gasbarro, Christina; Bello, Job M.; Bryan, Samuel A.; Lines, Amanda M.; Levitskaia, Tatiana G.

    2013-02-24T23:59:59.000Z

    Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source.

  10. Final report of the systems engineering technical advisory board for the Tank Waste Remediation Program

    SciTech Connect (OSTI)

    Baranowski, F.P.; Goodlett, C.B.; Beard, S.J.; Duckworth, J.P.; Schneider, A.; Zahn, L.L.

    1993-03-01T23:59:59.000Z

    The Tank Waste Remediation System (TWRS) is one segment of the environmental restoration program at the Hanford site. The scope is to retrieve the contents of both the single shell and double shell tanks and process the wastes into forms acceptable for long term storage and/or permanent disposal. The quantity of radioactive waste in tanks is significantly larger and substantially more complex in composition than the radioactive waste stored in tanks at other DOE sites. The waste is stored in 149 single shell tanks and 28 double shell tanks. The waste was produced over a period from the mid 1940s to the present. The single shell tanks have exceeded their design life and are experiencing failures. The oldest of the double shell tanks are approaching their design life. Spar double shell tank waste volume is limited. The priorities in the Board`s view are to manage safely the waste tank farms, accelerate emptying of waste tanks, provide spare tank capacity and assure a high degree of confidence in performance of the TWRS integrated program. At its present design capacity, the glass vitrification plant (HWVP) will require a period of about 15 years to empty the double shell tanks; the addition of the waste in single shell tanks adds another 100 years. There is an urgent need to initiate now a well focused and centralized development and engineering program on both larger glass melters and advanced separations processes that reduce radioactive constituents in the low-level waste (LLW). The Board presents its conclusions and has other suggestions for the management plan. The Board reviews planning schedules for accelerating the TWRS program.

  11. REGULATORY STRATEGIES TO MINIMIZE GENERATION OF REGULATED WASTES FROM CLEANUP, CONTINUED USE OR DECOMMISSIONING OF NUCLEAR FACILITIES CONTAMINATED WITH POLYCHLORINATED BIPHENYLS (PCBS) - 11198

    SciTech Connect (OSTI)

    Lowry, N.

    2010-11-05T23:59:59.000Z

    Disposal costs for liquid PCB radioactive waste are among the highest of any category of regulated waste. The high cost is driven by the fact that disposal options are extremely limited. Toxic Substances Control Act (TSCA) regulations require most liquids with PCBs at concentration of {ge} 50 parts-per-million to be disposed by incineration or equivalent destructive treatment. Disposal fees can be as high as $200 per gallon. This figure does not include packaging and the cost to transport the waste to the disposal facility, or the waste generator's labor costs for managing the waste prior to shipment. Minimizing the generation of liquid radioactive PCB waste is therefore a significant waste management challenge. PCB spill cleanups often generate large volumes of waste. That is because the removal of PCBs typically requires the liberal use of industrial solvents followed by a thorough rinsing process. In a nuclear facility, the cleanup process may be complicated by the presence of radiation and other occupational hazards. Building design and construction features, e.g., the presence of open grating or trenches, may also complicate cleanup. In addition to the technical challenges associated with spill cleanup, selection of the appropriate regulatory requirements and approach may be challenging. The TSCA regulations include three different sections relating to the cleanup of PCB contamination or spills. EPA has also promulgated a separate guidance policy for fresh PCB spills that is published as Subpart G of 40 CFR 761 although it is not an actual regulation. Applicability is based on the circumstances of each contamination event or situation. Other laws or regulations may also apply. Identification of the allowable regulatory options is important. Effective communication with stakeholders, particularly regulators, is just as important. Depending on the regulatory path that is taken, cleanup may necessitate the generation of large quantities of regulated waste. Allowable options must be evaluated carefully in order to reduce compliance risks, protect personnel, limit potential negative impacts on facility operations, and minimize the generation of wastes subject to TSCA. This paper will identify critical factors in selecting the appropriate TSCA regulatory path in order to minimize the generation of radioactive PCB waste and reduce negative impacts to facilities. The importance of communicating pertinent technical issues with facility staff, regulatory personnel, and subsequently, the public, will be discussed. Key points will be illustrated by examples from five former production reactors at the DOE Savannah River Site. In these reactors a polyurethane sealant was used to seal piping penetrations in the biological shield walls. During the intense neutron bombardment that occurred during reactor operation, the sealant broke down into a thick, viscous material that seeped out of the piping penetrations over adjacent equipment and walls. Some of the walls were painted with a PCB product. PCBs from the paint migrated into the degraded sealant, creating PCB 'spill areas' in some of these facilities. The regulatory cleanup approach selected for each facility was based on its operational status, e.g., active, inactive or undergoing decommissioning. The selected strategies served to greatly minimize the generation of radioactive liquid PCB waste. It is expected that this information would be useful to other DOE sites, DOD facilities, and commercial nuclear facilities constructed prior to the 1979 TSCA ban on most manufacturing and uses of PCBs.

  12. Chemistry, radiation, and interfaces in suspensions of nuclear waste simulants

    SciTech Connect (OSTI)

    Meisel, D.; Cook, A.; Camaioni, D.; Orlando, T.

    1997-08-01T23:59:59.000Z

    We focus in this report on three issues that are of central importance in the management of radioactive high-level liquid waste (HLLW). We show that the only reducing radical that lasts longer than a few ps in typical HLLW, and is capable of generating hydrogen, is NO{sub 3}{sup 2-}. We measured the lifetime of this species across the whole pH range (3 {le} pH {le} 14) and found it to be shorter than {approx}15 {mu}s, before is dissociates to give strongly oxidizing NO{sub 2} radicals. We found that it reacts with many proton donors (H{sup +} phosphate, borate, NH{sub 4}{sup +}, amines) in a reaction that is not merely an acid-base equilibrium reaction. Using high-level ab initio calculations we estimate its redox potential and pK{sub a}. We have developed methodologies to study the degradation of organic additives to the HLLW (to CO{sub 2} or carbonate) by NO{sub 2}. Relative rates of degradation of several complexants were determined using competition kinetics and {sup 13}C NMR and proton NMR detection techniques. Direct absorption of the radiation (low-energy electrons as well as photons above the ionization threshold) in NaNO{sub 3} single crystal at the solid/vacuum interface led to production of NO, O, and O- as the major products.

  13. MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES

    SciTech Connect (OSTI)

    Charles W. Solbrig; Kenneth J. Bateman

    2010-11-01T23:59:59.000Z

    The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn’t cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, “the length deficit,” produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

  14. Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

    SciTech Connect (OSTI)

    Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

    1995-10-20T23:59:59.000Z

    The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.

  15. Waste minimization handbook, Volume 1

    SciTech Connect (OSTI)

    Boing, L.E.; Coffey, M.J.

    1995-12-01T23:59:59.000Z

    This technical guide presents various methods used by industry to minimize low-level radioactive waste (LLW) generated during decommissioning and decontamination (D and D) activities. Such activities generate significant amounts of LLW during their operations. Waste minimization refers to any measure, procedure, or technique that reduces the amount of waste generated during a specific operation or project. Preventive waste minimization techniques implemented when a project is initiated can significantly reduce waste. Techniques implemented during decontamination activities reduce the cost of decommissioning. The application of waste minimization techniques is not limited to D and D activities; it is also useful during any phase of a facility`s life cycle. This compendium will be supplemented with a second volume of abstracts of hundreds of papers related to minimizing low-level nuclear waste. This second volume is expected to be released in late 1996.

  16. Addendum to the Calcined Waste Storage at the Idaho Nuclear Technology Center

    SciTech Connect (OSTI)

    M. D. Staiger; Michael Swenson; T. R. Thomas

    2004-05-01T23:59:59.000Z

    This report is an addendum to the report Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center, INEEL/EXT-98-00455 Rev. 1, June 2003. The original report provided a summary description of the Calcined Solids Storage Facilities (CSSFs). It also contained dozens of pages of detailed data tables documenting the volume and composition (chemical content and radionuclide activity) of the calcine stored in the CSSFs and the liquid waste from which the calcine was derived. This addendum report compiles the calcine composition data from the original report. It presents the compiled data in a graphical format with units (weight percent, curies per cubic meter, and nanocuries per gram) that are commonly used in regulatory and waste acceptance criteria documents. The compiled data are easier to use and understand when comparing the composition of the calcine with potential regulatory or waste acceptance criteria. This addendum report also provides detailed explanations for the large variability in the calcine composition among the CSSFs. The calcine composition varies as a result of reprocessing different types of fuel that had different cladding materials. Different chemicals were used to dissolve the various types of fuel, extract the uranium, and calcine the resulting waste. This resulted in calcine with variable compositions. This addendum report also identifies a few trace chemicals and radionuclides for which the accuracy of the amounts estimated to be in the calcine could be improved by making adjustments to the assumptions and methods used in making the estimates.

  17. Application of Direct Assessment Approaches and Methodologies to Cathodically Protected Nuclear Waste Transfer Lines

    SciTech Connect (OSTI)

    Dahl, Megan M. [ARES Corporation, Richland, WA (United States); Pikas, Joseph [Schiff Associates, Sugar Land TX (United States); Edgemon, Glenn L. [ARES Corporation, Richland, WA (United States); Philo, Sarah [ARES Corporation, Richland, WA (United States)

    2013-01-22T23:59:59.000Z

    The U.S. Department of Energy's (DOE) Hanford Site is responsible for the safe storage, retrieval, treatment, and disposal of approximately 54 million gallons (204 million liters) of radioactive waste generated since the site's inception in 1943. Today, the major structures involved in waste management at Hanford include 149 carbon steel single-shell tanks, 28 carbon-steel double-shell tanks, plus a network of buried metallic transfer lines and ancillary systems (pits, vaults, catch tanks, etc.) required to store, retrieve, and transfer waste within the tank farm system. Many of the waste management systems at Hanford are still in use today. In response to uncertainties regarding the structural integrity of these systems,' an independent, comprehensive integrity assessment of the Hanford Site piping system was performed. It was found that regulators do not require the cathodically protected pipelines located within the Hanford Site to be assessed by External Corrosion Direct Assessment (ECDA) or any other method used to ensure integrity. However, a case study is presented discussing the application of the direct assessment process on pipelines in such a nuclear environment. Assessment methodology and assessment results are contained herein. An approach is described for the monitoring, integration of outside data, and analysis of this information in order to identify whether coating deterioration accompanied by external corrosion is a threat for these waste transfer lines.

  18. Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 3: Environmental remediation and environmental management issues

    SciTech Connect (OSTI)

    Baschwitz, R.; Kohout, R.; Marek, J.; Richter, P.I.; Slate, S.C. [eds.

    1993-12-31T23:59:59.000Z

    This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Papers are divided into the following sections: Low/Intermediate level waste disposal from an international viewpoint; Solid waste volume reduction, treatment and packaging experience; Design of integrated systems for management of nuclear wastes; Mixed waste (hazardous and radioactive) treatment and disposal; Advanced low/intermediate level waste conditioning technologies including incineration; National programs for low/intermediate waste management; Low/Intermediate waste characterization, assay, and tracking systems; Disposal site characterization and performance assessment; Radioactive waste management and practices in developing countries; Waste management from unconventional (e.g. VVER) nuclear power reactors; Waste minimization, avoidance and recycling in nuclear power plants; Liquid waste treatment processes and experience; Low/Intermediate waste storage facilities--design and experience; Low/Intermediate waste forms and acceptance criteria for disposal; Management of non-standard or accident waste; and Quality assurance and control in nuclear waste management. Individual papers have been processed separately for inclusion in the appropriate data bases.

  19. Technical Guidance

    Broader source: Energy.gov [DOE]

    The Office of Technical Guidance, within the Office of Health, Safety and Security develops and issues Government-wide and Department-wide technical guidance to ensure that classified nuclear...

  20. Comparison of waste emplacement configurations for a nuclear waste respository in tuff. IV. Thermo-hydrological analysis

    SciTech Connect (OSTI)

    Mondy, L.A.; Wilson, R.K.; Bixler, N.E.

    1983-08-01T23:59:59.000Z

    This report summarizes the results of a hydrological analysis of two emplacement schemes being considered for the storage of commercial high-level nuclear waste at the Nevada Test Site. The analysis is two-dimensional, considers the flow of water in partially saturated tuff (the Topopah Springs member of the Paintbrush tuff in Yucca Mountain) and includes the effects of the heat source (waste canisters) on that flow. The results include measures of the heat flux entering the access and emplacement drifts, measures of the flow rates near the canisters and a comparison of the temperature fields. It was necessary in the analysis to approximate the boundary conditions at the walls of the access and emplacement drifts in order to simulate the ventilation process. As a result the analysis was done for several cases which were expected to bracket the actual situation. A discussion of this problem is also included in the report. It should be noted that these results are intended as a means of comparing emplacement schemes, not as a performance assessment. 72 figures, 6 tables.

  1. Electrical and electronic subsystems of a nuclear waste tank annulus inspection system

    SciTech Connect (OSTI)

    Evenson, R.J.

    1981-06-01T23:59:59.000Z

    The nuclear waste tank annulus inspection system is designed specifically for use at the Nuclear Regulatory Commission's Nuclear Fuel Services Facility at West Valley, New York. This system sends a television and photographic camera into the space between the walls of a double-shell nuclear waste tank to obtain images of the inner and outer walls at precisely known locations. The system is capable of inspecting a wall section 14 ft wide by 27 ft high. Due to the high temperature and radiation of the annulus environment, the operating life for the inspection device is uncertain, but is expected to be at least 100 h, with 1000 R/h at 82/sup 0/C. The film camera is shielded with 1/2 in. of lead to minimize radiation fogging of the film during a 25-min picture taking excursion. The operation of the inspection system is semiautomated with remote manual prepositioning of the camera, followed by a computer controlled wall scan. This apparatus is currently set up to take an array of contiguous pictures, but is adaptable to other modes of operation.

  2. TECHNICAL ASSESSMENT OF FRACTIONAL CRYSTALLIZATION FOR TANK WASTE PRETREATMENT AT THE DOE HANFORD SITE

    SciTech Connect (OSTI)

    HAMILTON, D.W.

    2006-01-03T23:59:59.000Z

    Radioactive wastes from one hundred seventy-seven underground storage tanks in the 200 Area of the Department of Energy (DOE) Hanford Site in Washington State will be retrieved, treated and stored either on site or at an approved off-site repository. DOE is currently planning to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions, which would be treated and permanently disposed in separate facilities. A significant volume of the wastes in the Hanford tanks is currently classified as medium Curie waste, which will require separation and treatment at the Waste Treatment Plant (WTP). Because of the specific challenges associated with treating this waste stream, DOE EM-21 funded a project to investigate the feasibility of using fractional crystallization as a supplemental pretreatment technology. The two process requirements for fractional crystallization to be successfully applied to Hanford waste include: (1) evaporation of water from the aqueous solution to enrich the activity of soluble {sup 137}Cs, resulting in a higher activity stream to be sent to the WTP, and (2) separation of the crystalline salts that are enriched in sodium, carbonate, sulfate, and phosphate and sufficiently depleted in {sup 137}Cs, to produce a second stream to be sent to Bulk Vitrification. Phase I of this project has just been completed by COGEMA/Georgia Institute of Technology. The purpose of this report is to document an independent expert review of the Phase I results with recommendations for future testing. A team of experts with significant experience at both the Hanford and Savannah River Sites was convened to conduct the review at Richland, Washington the week of November 14, 2005.

  3. Coolside waste management research. Annual technical progress report, October 1992--September 1993

    SciTech Connect (OSTI)

    Not Available

    1993-10-01T23:59:59.000Z

    The composition (major, minor and trace elements) of approximately 400 Coolside waste samples form the Edgewater Coolside demonstration and CONSOL pilot plant tests are summarized and tabulated in this report. The composition changes in the waste samples collected during the Edgewater Coolside demonstration can be correlated with the processing variables and operation conditions. A study was conducted that focused on the formation mechanism of ettringite crystals in FGD waste materials since they were observed to form in Coolside, FBC, and AFBC derived wastes. The degree of swelling in FGD-derived waste can be correlated to the amounts of ettringite crystals present. Reactions controlling the formation, composition and disintegration of ettringite are critical in determining the overall stability and strength of cements and concretes derived from dry-flue gas desulfurization wastes. Since these wastes consist of fly ash, along with, CaSO{sub 3}, and CaSO{sub 4} and unreacted Ca(OH){sub 2}, the sulfites and sulfates react with the Ca(OH){sub 2} along with the glassy aluminosilicates in the fly ash to form calcium sulfo-aluminate minerals. Ettringite, the most important of these, is the main contributor to the compressive strength development of the FGD waste mixtures. However, excessive ettringite formation causes swelling which often leads to destructive crack formation. It has been shown that the quantity of ettringite formed and compressive strength of the FGD waste mixtures reach a maximum until ettringite begins to disintegrate. Because the formation mechanisms of ettringite are not entirely understood, swelling in FGD derived products is difficult to predict.

  4. Savannah River Site - Salt Waste Processing Facility: Briefing on the Salt Waste Processing Facility Independent Technical Review

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmentalEfficiency | Department ofEnergySaraKSALT WASTE

  5. Using electrochemical separation to reduce the volume of high-level nuclear waste

    SciTech Connect (OSTI)

    Slater, S.A.; Gay, E.C.

    1998-07-01T23:59:59.000Z

    Argonne National Laboratory (ANL) has developed an electrochemical separation technique called electrorefining that will treat a variety of metallic spent nuclear fuel and reduce the volume of high-level nuclear waste that requires disposal. As part of that effort, ANL has developed a high throughput electrorefiner (HTER) that has a transport rate approximately three times faster than electrorefiners previously developed at ANL. This higher rate is due to the higher electrode surface area, a shorter transport path, and more efficient mixing, which leads to smaller boundary layers about the electrodes. This higher throughput makes electrorefining an attractive option in treating Department of Energy spent nuclear fuels. Experiments have been done to characterize the HTER, and a simulant metallic fuel has been successfully treated. The HTER design and experimental results is discussed.

  6. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    SciTech Connect (OSTI)

    Hill, Thomas J

    2005-09-01T23:59:59.000Z

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

  7. Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

    SciTech Connect (OSTI)

    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.

    1990-09-01T23:59:59.000Z

    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.

  8. Treatment technologies for hazardous ashes generated from possible incineration of navy waste. Technical note

    SciTech Connect (OSTI)

    Torres, T.

    1990-10-01T23:59:59.000Z

    The Navy recognizes that thermal treatment of Navy hazardous wastes (HW) should, under the terms of the Resource Conservation and Recovery Act of 1976, be avoided. Combustion waste disposal may nonetheless become unavoidable in certain cases, even after all possible process enhancements that avoid HW production are implemented. Even then, some toxic constituents that may be present in the waste will not be destroyed by incineration and will persist in the ash residue produced by incineration. Such incinerator ashes will have to be disposed of in HW landfills. The Navy is thus evaluating methods of treatment of such ash to remove or immobilize the toxic constituents that persist following incineration in order to render the waste treatment residue nonhazardous. Appropriate technology identified in this work can be applied to ash produced by HW combuster operated by the Navy, if any, or be required for ash produced by commercial generators handling Navy HWs.

  9. Reviewing PSA-based analyses to modify technical specifications at nuclear power plants

    SciTech Connect (OSTI)

    Samanta, P.K.; Martinez-Guridi, G. [Brookhaven National Lab., Upton, NY (United States); Vesely, W.E. [Science Applications International Corporation, Dublin, OH (United States)

    1995-12-01T23:59:59.000Z

    Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant`s probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed.

  10. Preliminary technical data summary No. 3 for the Defense Waste Processing Facility

    SciTech Connect (OSTI)

    Landon, L.F. (comp.)

    1980-05-01T23:59:59.000Z

    This document presents an update on the best information presently available for the purpose of establishing the basis for the design of a Defense Waste Processing Facility. Objective of this project is to provide a facility to fix the radionuclides present in Savannah River Plant (SRP) high-level liquid waste in a high-integrity form (glass). Flowsheets and material balances reflect the alternate CAB case including the incorporation of low-level supernate in concrete. (DLC)

  11. RH-TRU Waste Shipments from Battelle Columbus Laboratories to the Hanford Nuclear Facility for Interim Storage

    SciTech Connect (OSTI)

    Eide, J.; Baillieul, T. A.; Biedscheid, J.; Forrester, T,; McMillan, B.; Shrader, T.; Richterich, L.

    2003-02-26T23:59:59.000Z

    Battelle Columbus Laboratories (BCL), located in Columbus, Ohio, must complete decontamination and decommissioning (D&D) activities for nuclear research buildings and grounds by 2006, as directed by Congress. Most of the resulting waste (approximately 27 cubic meters [m3]) is remote-handled (RH) transuranic (TRU) waste destined for disposal at the Waste Isolation Pilot Plant (WIPP). The BCL, under a contract to the U.S. Department of Energy (DOE) Ohio Field Office, has initiated a plan to ship the TRU waste to the DOE Hanford Nuclear Facility (Hanford) for interim storage pending the authorization of WIPP for the permanent disposal of RH-TRU waste. The first of the BCL RH-TRU waste shipments was successfully completed on December 18, 2002. This BCL shipment of one fully loaded 10-160B Cask was the first shipment of RH-TRU waste in several years. Its successful completion required a complex effort entailing coordination between different contractors and federal agencies to establish necessary supporting agreements. This paper discusses the agreements and funding mechanisms used in support of the BCL shipments of TRU waste to Hanford for interim storage. In addition, this paper presents a summary of the efforts completed to demonstrate the effectiveness of the 10-160B Cask system. Lessons learned during this process are discussed and may be applicable to other TRU waste site shipment plans.

  12. Plutonium Equivalent Inventory for Belowground Radioactive Waste at the Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

    SciTech Connect (OSTI)

    French, Sean B. [Los Alamos National Laboratory; Shuman, Rob [WPS: WASTE PROJECTS AND SERVICES

    2012-04-18T23:59:59.000Z

    The Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Many aspects of the management of this waste are conducted at Technical Area 54 (TA-54); Area G plays a key role in these management activities as the Laboratory's only disposal facility for low-level radioactive waste (LLW). Furthermore, Area G serves as a staging area for transuranic (TRU) waste that will be shipped to the Waste Isolation Pilot Plant for disposal. A portion of this TRU waste is retrievably stored in pits, trenches, and shafts. The radioactive waste disposed of or stored at Area G poses potential short- and long-term risks to workers at the disposal facility and to members of the public. These risks are directly proportional to the radionuclide inventories in the waste. The Area G performance assessment and composite analysis (LANL, 2008a) project long-term risks to members of the public; short-term risks to workers and members of the public, such as those posed by accidents, are addressed by the Area G Documented Safety Analysis (LANL, 2011a). The Documented Safety Analysis uses an inventory expressed in terms of plutonium-equivalent curies, referred to as the PE-Ci inventory, to estimate these risks. The Technical Safety Requirements for Technical Area 54, Area G (LANL, 2011b) establishes a belowground radioactive material limit that ensures the cumulative projected inventory authorized for the Area G site is not exceeded. The total belowground radioactive waste inventory limit established for Area G is 110,000 PE-Ci. The PE-Ci inventory is updated annually; this report presents the inventory prepared for 2011. The approach used to estimate the inventory is described in Section 2. The results of the analysis are presented in Section 3.

  13. Uranium, thorium and trace elements in geologic occurrences as analogues of nuclear waste repository conditions

    SciTech Connect (OSTI)

    Wollenberg, H.A.; Brookins, D.G.; Cohen, L.H.; Flexser, S.; Abashian, M.; Murphy, M.; Williams, A.E.

    1984-05-01T23:59:59.000Z

    Contact zones between intrusive rocks and tuff, basalt, salt and granitic rock were investigated as possible analogues of nuclear waste repository conditions. Results of detailed studies of contacts between quartz monzonite of Laramide age, intrusive into Precambrian gneiss, and a Tertiary monzonite-tuff contact zone indicate that uranium, thorium and other trace elements have not migrated significantly from the more radioactive instrusives into the country rock. Similar observations resulted from preliminary investigations of a rhyodacite dike cutting basalt of the Columbia River plateau and a kimberlitic dike cutting bedded salt of the Salina basin. This lack of radionuclide migration occurred in hydrologic and thermal conditions comparable to, or more severe than those expected in nuclear waste repository environments and over time periods of the order of concern for waste repositories. Attention is now directed to investigation of active hydrothermal systems in candidate repository rock types, and in this regard a preliminary set of samples has been obtained from a core hole intersecting basalt underlying the Newberry caldera, Oregon, where temperatures presently range from 100 to 265{sup 0}C. Results of mineralogical and geochemical investigations of this core should indicate the alteration mineralogy and behavior of radioelements in conditions analogous to those in the near field of a repository in basalt.

  14. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    SciTech Connect (OSTI)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01T23:59:59.000Z

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  15. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model - 13413

    SciTech Connect (OSTI)

    Djokic, Denia [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States)] [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States); Piet, Steven J.; Pincock, Layne F.; Soelberg, Nick R. [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)] [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)

    2013-07-01T23:59:59.000Z

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system, and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity. (authors)

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

    SciTech Connect (OSTI)

    Pankov, Alexey S.; Ojovan, Michael I. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield, S1 3JD (United Kingdom); Batyukhnova, Olga G. [International Education Training Centre, SUE SIA 'Radon', The 7-th Rostovsky Lane 2/14, Moscow, 119121 (Russian Federation); Lee, William E. [Department of Materials, Imperial College London, South Kensington Campus, Exhibition Road, London, SW7 2AZ (United Kingdom)

    2007-07-01T23:59:59.000Z

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

  17. Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

    SciTech Connect (OSTI)

    Cottrell, W.B.; Passiakos, M.

    1982-06-01T23:59:59.000Z

    This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

  18. ANL technical support program for DOE environmental restoration and waste management. Annual report, October 1993--September 1994

    SciTech Connect (OSTI)

    Bates, J.K.; Brown, N.R.; Buck, E.C. [and others

    1995-06-01T23:59:59.000Z

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1994 on the following tasks: (1) Critical Reviews of important parameters that affect the reactivity of glass in an unsaturated environment are being prepared. (2) A series of tests is ongoing to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. (3) The effect of radiation upon the durability of waste glasses at a high SA/V ratio and a high gas-to-liquid volume ratio has been assessed. (4) A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SA/V ratios. Such differences in the SA/V ratio may significantly affect glass durability. At long-term periods and high SA/V ratios, acceleration in glass reaction has been observed. (5) Tests were initiated on West Valley Reference 6 (WV6) glass and on the Environmental Assessment (EA) glass. (6) Tests with the actinide-doped West Valley glass ATM-10 have been in progress for over seven years as a part of work for the Yucca Mountain Site Characterization Project (YMP). (7) Analytical electron microscopy (AEM) is being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Also, colloids from the leach solutions are being studied using AEM.

  19. Recommendation by the Secretary of Energy Regarding the Suitability of the Yucca Mountain Site for a Repository Under the Nuclear Waste Policy Act of 1982

    Broader source: Energy.gov [DOE]

    Recommendation by the Secretary of Energy Regarding the Suitability of the Yucca Mountain Site for a Repository Under the Nuclear Waste Policy Act of 1982

  20. Hanford Site Secondary Waste Roadmap

    SciTech Connect (OSTI)

    Westsik, Joseph H.

    2009-01-29T23:59:59.000Z

    Summary The U.S. Department of Energy (DOE) is making plans to dispose of 54 million gallons of radioactive tank wastes at the Hanford Site near Richland, Washington. The high-level wastes and low-activity wastes will be vitrified and placed in permanent disposal sites. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents, and these need to be processed and disposed of also. The Department of Energy Office of Waste Processing sponsored a meeting to develop a roadmap to outline the steps necessary to design the secondary waste forms. Representatives from DOE, the U.S. Environmental Protection Agency, the Washington State Department of Ecology, the Oregon Department of Energy, Nuclear Regulatory Commission, technical experts from the DOE national laboratories, academia, and private consultants convened in Richland, Washington, during the week of July 21-23, 2008, to participate in a workshop to identify the risks and uncertainties associated with the treatment and disposal of the secondary wastes and to develop a roadmap for addressing those risks and uncertainties. This report describes the results of the roadmap meeting in Richland. Processing of the tank wastes will generate secondary wastes, including routine solid wastes and liquid process effluents. The secondary waste roadmap workshop focused on the waste streams that contained the largest fractions of the 129I and 99Tc that the Integrated Disposal Facility risk assessment analyses were showing to have the largest contribution to the estimated IDF disposal impacts to groundwater. Thus, the roadmapping effort was to focus on the scrubber/off-gas treatment liquids with 99Tc to be sent to the Effluent Treatment Facility for treatment and solidification and the silver mordenite and carbon beds with the captured 129I to be packaged and sent to the IDF. At the highest level, the secondary waste roadmap includes elements addressing regulatory and performance requirements, waste composition, preliminary waste form screening, waste form development, process design and support, and validation. The regulatory and performance requirements activity will provide the secondary waste-form performance requirements. The waste-composition activity will provide workable ranges of secondary waste compositions and formulations for simulants and surrogates. Preliminary waste form screening will identify candidate waste forms for immobilizing the secondary wastes. The waste form development activity will mature the waste forms, leading to a selected waste form(s) with a defensible understanding of the long-term release rate and input into the critical decision process for a secondary waste treatment process/facility. The process and design support activity will provide a reliable process flowsheet and input to support a robust facility design. The validation effort will confirm that the selected waste form meets regulatory requirements. The final outcome of the implementation of the secondary waste roadmap is the compliant, effective, timely, and cost-effective disposal of the secondary wastes. The work necessary to address the programmatic, regulatory, and technical risks and uncertainties identified through the Secondary Waste Roadmap Workshop are assembled into several program needs elements. Programmatic/Regulatory needs include: • Select and deploy Hanford tank waste supplemental treatment technology • Provide treatment capability for secondary waste streams from tank waste treatment • Develop consensus on secondary waste form acceptance. Technology needs include: • Define secondary waste composition ranges and uncertainties • Identify and develop waste forms for secondary waste immobilization and disposal • Develop test methods to characterize secondary waste form performance. Details for each of these program elements are provided.

  1. Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

    SciTech Connect (OSTI)

    Saurwein, John

    2011-07-15T23:59:59.000Z

    This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

  2. Summary Notes from 28 May 2008 Generic Technical Issue Discussion on Estimating Waste Inventory and Waste Tank Characterization

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssues DOE's NuclearSpurringSteamDepartment4SuitlandOctober 12, 2011

  3. Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste

    SciTech Connect (OSTI)

    Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

    1993-09-01T23:59:59.000Z

    An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

  4. Nevada Nuclear Waste Storage Investigations: Exploratory Shaft Facility fluids and materials evaluation

    SciTech Connect (OSTI)

    West, K.A.

    1988-11-01T23:59:59.000Z

    The objective of this study was to determine if any fluids or materials used in the Exploratory Shaft Facility (ESF) of Yucca Mountain will make the mountain unsuitable for future construction of a nuclear waste repository. Yucca Mountain, an area on and adjacent to the Nevada Test Site in southern Nevada, USA, is a candidate site for permanent disposal of high-level radioactive waste from commercial nuclear power and defense nuclear activities. To properly characterize Yucca Mountain, it will be necessary to construct an underground test facility, in which in situ site characterization tests can be conducted. The candidate repository horizon at Yucca Mountain, however, could potentially be compromised by fluids and materials used in the site characterization tests. To minimize this possibility, Los Alamos National Laboratory was directed to evaluate the kinds of fluids and materials that will be used and their potential impacts on the site. A secondary objective was to identify fluids and materials, if any, that should be prohibited from, or controlled in, the underground. 56 refs., 19 figs., 11 tabs.

  5. Emptying of the Storage for Solid Radioactive Waste in the Greifswald Nuclear Power Plant

    SciTech Connect (OSTI)

    Hartmann, B.; Fischer, J.

    2002-02-26T23:59:59.000Z

    On the Greifswald site, 8 WWER 440 reactor units are located and also several facilities to handle fuel and radwaste. After the reunification of Germany, the final decision was taken to decommission all these Russian designed reactors. Thus, EWN is faced with a major decommissioning project in the field of nuclear power stations. One of the major tasks before the dismantling of the plant is the complete disposal of the operational waste. Among other facilities, a store for solid radioactive waste is located on the site, which has been filled over 17 years of operation of units 1 to 4. The paper presents the disposal technology development and results achieved. This activity is the first project in the operational history of the Russian type serial reactor line WWER-440.

  6. Solid Waste Projection Model: Database (Version 1.4). Technical reference manual

    SciTech Connect (OSTI)

    Blackburn, C.; Cillan, T.

    1993-09-01T23:59:59.000Z

    The Solid Waste Projection Model (SWPM) system is an analytical tool developed by Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company (WHC). The SWPM system provides a modeling and analysis environment that supports decisions in the process of evaluating various solid waste management alternatives. This document, one of a series describing the SWPM system, contains detailed information regarding the software and data structures utilized in developing the SWPM Version 1.4 Database. This document is intended for use by experienced database specialists and supports database maintenance, utility development, and database enhancement. Those interested in using the SWPM database should refer to the SWPM Database User`s Guide. This document is available from the PNL Task M Project Manager (D. L. Stiles, 509-372-4358), the PNL Task L Project Manager (L. L. Armacost, 509-372-4304), the WHC Restoration Projects Section Manager (509-372-1443), or the WHC Waste Characterization Manager (509-372-1193).

  7. Solid Waste Projection Model: Database (Version 1.3). Technical reference manual

    SciTech Connect (OSTI)

    Blackburn, C.L.

    1991-11-01T23:59:59.000Z

    The Solid Waste Projection Model (SWPM) system is an analytical tool developed by Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company (WHC). The SWPM system provides a modeling and analysis environment that supports decisions in the process of evaluating various solid waste management alternatives. This document, one of a series describing the SWPM system, contains detailed information regarding the software and data structures utilized in developing the SWPM Version 1.3 Database. This document is intended for use by experienced database specialists and supports database maintenance, utility development, and database enhancement.

  8. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

    SciTech Connect (OSTI)

    Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

    2011-02-01T23:59:59.000Z

    This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

  9. Technical Note: Updated durability/composition relationships for Hanford high-level waste glasses

    SciTech Connect (OSTI)

    Piepel, G.F.; Hartley, S.A.; Redgate, P.E.

    1996-03-01T23:59:59.000Z

    This technical note presents empirical models developed in FYI 995 to predict durability as functions of glass composition. Models are presented for normalized releases of B, Li, Na, and Si from the 7-day Product Consistency Test (PCT) applied to quenched and canister centerline cooled (CCC) glasses as well as from the 28-day Materials Characterization Center-1 (MCC-1) test applied to quenched glasses. Models are presented for Composition Variation Study (CVS) data from low temperature melter (LTM) studies (Hrma, Piepel, et al. 1994) and high temperature melter (HTM) studies (Vienna et al. 1995). The data used for modeling in this technical note are listed in Appendix A.

  10. A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles

    SciTech Connect (OSTI)

    Peters, M.T. [Applied Science and Technology, Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL, 60439 (United States); Ewing, R.C. [Department of Geological Sciences, The University of Michigan, 2534 C.C. Little Bldg., 1100 N. University, Ann Arbor, MI, 48109-1005 (United States)

    2007-07-01T23:59:59.000Z

    There are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: a) SNF dissolution mechanisms and rates; b) formation and properties of U{sup 6+}- secondary phases; c) waste form-waste package interactions in the near-field; and d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 10{sup 5} years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms 'tailored' to specific geologic settings. (authors)

  11. Nuclear waste processing based on FOOF and KrF sub 2

    SciTech Connect (OSTI)

    Kim, K.C.; Blum, T.W.

    1991-01-01T23:59:59.000Z

    A class of oxygen fluoride compounds and krypton difluoride show great promise in recovering and purifying plutonium and other actinides from nuclear waste and residues. Since 1983, a significant effort has been expended in three main areas of research; one area is the characterization of the reactive species and the fluorination reaction chemistry with various actinides. The second area is to develop efficient production methods for the active chemicals. The third area is actually in demonstrating application of these fluorinating agents to actinides recovery and purification. Substantial progress has been made in all three areas and some of the highlights in their research effort is discussed. 11 refs., 2 figs., 2 tabs.

  12. Technical and Political Assessment of Peaceful Nuclear Power Program Prospects in North Africa and the Middle East

    SciTech Connect (OSTI)

    Windsor, Lindsay K.; Kessler, Carol E.

    2007-09-11T23:59:59.000Z

    An exceptional number of Middle Eastern and North African nations have recently expressed interest in developing nuclear energy for peaceful purposes. Many of these countries have explored nuclear research in limited ways in the past, but the current focused interest and application of resources towards developing nuclear-generated electricity and nuclear-powered desalination plants is unprecedented. Consequently, questions arise in response to this emerging trend: What instigated this interest? To what end(s) will a nuclear program be applied? Does the country have adequate technical, political, legislative, nonproliferation, and safety infrastructure required for the capability desired? If so, what are the next steps for a country in preparation for a future nuclear program? And if not, what collaboration efforts are possible with the United States or others? This report provides information on the capabilities and interests of 13 countries in the region in nuclear energy programs in light of safety, nonproliferation and security concerns. It also provides information useful for determining potential for offering technical collaboration, financial aid, and/or political support.

  13. Preliminary Technical and Legal Evaluation of Disposing of Nonhazardous Oil Field Waste into Salt Caverns

    SciTech Connect (OSTI)

    Ayers, Robert C.; Caudle, Dan; Elcock, Deborah; Raivel, Mary; Veil, John; and Grunewald, Ben

    1999-01-21T23:59:59.000Z

    This report presents an initial evaluation of the suitability, feasibility, and legality of using salt caverns for disposal of nonhazardous oil field wastes. Given the preliminary and general nature of this report, we recognize that some of our findings and conclusions maybe speculative and subject to change upon further research on this topic.

  14. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    SciTech Connect (OSTI)

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01T23:59:59.000Z

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

  15. Preliminary data from an instantaneous profile test conducted near the Mixed Waste Landfill, Technical Area 3, Sandia National Laboratories/New Mexico

    SciTech Connect (OSTI)

    Bayliss, S.C. [DanShar, Inc., Bosque Farms, NM (United States); Goering, T.J.; McVey, M.D. [GRAM, Inc., Albuquerque, NM (United States); Strong, W.R.; Peace, J.L. [Sandia National Labs., Albuquerque, NM (United States). Environmental Restoration Project

    1996-04-01T23:59:59.000Z

    This paper presents data from an instantaneous profile test conducted near the Sandia National Laboratories/New Mexico Mixed Waste Landfill in Technical Area 3. The test was performed from December 1993 through 1995 as part of the environmental Restoration Project`s Phase 2 RCRA Facility Investigation of the Mixed Waste Landfill. The purpose of the test was to measure the unsaturated hydraulic properties of soils near the Mixed Waste Landfill. The instantaneous profile test and instrumentation are described, and the pressure and moisture content data from the test are presented. These data may be useful for understanding the unsaturated hydraulic properties of soils in Technical Area 3 and for model validation, verification, and calibration.

  16. Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.

    SciTech Connect (OSTI)

    Zapol, Peter (Argonne National Laboratory, Argonne, IL); Bourg, Ian (Lawrence Berkeley National Laboratories, Berkeley, CA); Criscenti, Louise Jacqueline; Steefel, Carl I. (Lawrence Berkeley National Laboratories, Berkeley, CA); Schultz, Peter Andrew

    2011-10-01T23:59:59.000Z

    This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

  17. The National Nuclear Laboratory's Approach to Processing Mixed Wastes and Residues - 13080

    SciTech Connect (OSTI)

    Greenwood, Howard; Docrat, Tahera; Allinson, Sarah J.; Coppersthwaite, Duncan P.; Sultan, Ruqayyah; May, Sarah [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)] [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)

    2013-07-01T23:59:59.000Z

    The National Nuclear Laboratory (NNL) treats a wide variety of materials produced as by-products of the nuclear fuel cycle, mostly from uranium purification and fuel manufacture but also including materials from uranium enrichment and from the decommissioning of obsolete plants. In the context of this paper, treatment is defined as recovery of uranium or other activity from residues, the recycle of uranium to the fuel cycle or preparation for long term storage and the final disposal or discharge to the environment of the remainder of the material. NNL's systematic but flexible approach to residue assessment and treatment is described in this paper. The approach typically comprises up to five main phases. The benefits of a systematic approach to waste and residue assessments and processing are described in this paper with examples used to illustrate each phase of work. Benefits include early identification of processing routes or processing issues and the avoidance of investment in inappropriate and costly plant or processes. (authors)

  18. International Nuclear Energy Research Initiative Development of Computational Models for Pyrochemical Electrorefiners of Nuclear Waste Transmutation Systems

    SciTech Connect (OSTI)

    M.F. Simpson; K.-R. Kim

    2010-12-01T23:59:59.000Z

    In support of closing the nuclear fuel cycle using non-aqueous separations technology, this project aims to develop computational models of electrorefiners based on fundamental chemical and physical processes. Spent driver fuel from Experimental Breeder Reactor-II (EBR-II) is currently being electrorefined in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory (INL). And Korea Atomic Energy Research Institute (KAERI) is developing electrorefining technology for future application to spent fuel treatment and management in the Republic of Korea (ROK). Electrorefining is a critical component of pyroprocessing, a non-aqueous chemical process which separates spent fuel into four streams: (1) uranium metal, (2) U/TRU metal, (3) metallic high-level waste containing cladding hulls and noble metal fission products, and (4) ceramic high-level waste containing sodium and active metal fission products. Having rigorous yet flexible electrorefiner models will facilitate process optimization and assist in trouble-shooting as necessary. To attain such models, INL/UI has focused on approaches to develop a computationally-light and portable two-dimensional (2D) model, while KAERI/SNU has investigated approaches to develop a computationally intensive three-dimensional (3D) model for detailed and fine-tuned simulation.

  19. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect (OSTI)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18T23:59:59.000Z

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

  20. Preliminary technique assessment for nondestructive evaluation certification of the NNWSI [Nevada Nuclear Waste Storage Investigations] disposal container closure

    SciTech Connect (OSTI)

    Day, R.A.

    1988-12-31T23:59:59.000Z

    Under the direction of the Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM) program, the Nevada Nuclear Waste Storage Investigations (NNWSI) project is evaluating a candidate repository site at Yucca Mountain, Nevada, for permanent disposal of high-level nuclear waste. The Lawrence Livermore National Laboratory (LLNL), a participant in the NNWSI project, is developing waste package designs to meet the NRC requirements. One aspect of this waste package is the nondestructive testing of the final closure of the waste container. The container closure weld can best be nondestructively examined (NDE) by a combination of ultrasonics and liquid penetrants. This combination can be applied remotely and can meet stringent quality control requirements common to nuclear applications. Further development in remote systems and inspection will be required to meet anticipated requirements for flaw detection reliability and sensitivity. New research is not required but might reduce cost or inspection time. Ultrasonic and liquid penetrant methods can examine all closure methods currently being considered, which include fusion welding and inertial welding, among others. These NDE methods also have a history of application in high radiation environments and a well developed technology base for remote operation that can be used to reduce development and design costs. 43 refs., 23 figs., 3 tabs.

  1. Geologic and hydrologic investigations of a potential nuclear waste disposal site at Yucca Mountain, southern Nevada

    SciTech Connect (OSTI)

    Carr, M.D.; Yount, J.C. (eds.)

    1988-12-31T23:59:59.000Z

    Yucca Mountain in southern Nye County, Nevada, has been selected by the United States Department of Energy as one of three potential sites for the nation`s first high-level nuclear waste repository. Its deep water table, closed-basin ground-water flow, potentially favorable host rock, and sparse population have made the Yucca Mountain area a viable candidate during the search for a nuclear waste disposal site. Yucca Mountain, however, lies within the southern Great Basin, a region of known contemporary tectonism and young volcanic activity, and the characterization of tectonism and volcanism remains as a fundamental problem for the Yucca Mountain site. The United States Geological Survey has been conducting extensive studies to evaluate the geologic setting of Yucca Mountain, as well as the timing and rates of tectonic and volcanic activity in the region. A workshop was convened by the Geologic Survey in Denver, Colorado, on August 19, 20, and 21, 1985, to review the scientific progress and direction of these studies. Considerable debate resulted. This collection of papers represents the results of some of the studies presented at the workshop, but by no means covers all of the scientific results and viewpoints presented. Rather, the volume is meant to serve as a progress report on some of the studies within the Geological Survey`s continuing research program toward characterizing the tectonic framework of Yucca Mountain. Individual papers were processed separately for the data base.

  2. Development of permeable reactive barriers to prevent radionuclide migration from the nuclear waste repositories

    SciTech Connect (OSTI)

    Zakharova, E. [Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow (Russian Federation); Kalmykov, S.; Batuk, O. [Chemistry department of Lomonosov Moscow, State University, Moscow (Russian Federation); Kazakovskaya, T.; Shapovalov, V. [All-Russian Scientific Research Institute of Experimental Physics (Russian Federation); Haire, M.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2007-07-01T23:59:59.000Z

    This paper is focused on three possible materials for permeable reactive barriers (PRB): 1) depleted uranium oxide that is accumulated as a residual product of the natural uranium enrichment process, 2) zero-valent iron and, 3) the composite material based on montmorillonite clay modified with different anion exchangers. The main aim of permeable reactive barriers is to prevent release of radionuclides emerging from a repository waste package containing spent nuclear fuel to outside the control area of the nuclear waste repository sites. The most experimentally developed material is depleted uranium oxide. It can be used both as a component of radiation shielding and as an absorbent for migrating long-lived radionuclides (especially {sup 237}Np and {sup 99}Tc). Experiments demonstrate the high sorption properties of depleted uranium oxide towards Np and Tc both from deionized water and from solution that simulates Yucca Mountain. Zero-valent iron, and the composite based on montmorillonite clay, also seem to be very promising to use in a PRB. Nano-particles of zero-valent iron with high surface will reduce high valency Np and Tc to the tetravalent state and thus immobilize them due to the extremely low solubility of corresponding hydroxides. The composite based on montmorillonite clay modified with different anion exchangers will possess high sorption affinity towards anionic and cationic species. (authors)

  3. Summary of four release consequence analyses for hypothetical nuclear waste repositories in salt and granite

    SciTech Connect (OSTI)

    Cole, C.R.; Bond, F.W.

    1980-12-01T23:59:59.000Z

    Release consequence methology developed under the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) program has now been applied to four hypothetical repository sites. This paper summarizes the results of these four studies in order to demonstrate that the far-field methodology developed under the AEGIS program offers a practical approach to the post-closure safety assessment of nuclear waste repositories sited in deep continental geologic formations. The four studies are briefly described and compared according to the following general categories: physical description of the repository (size, inventory, emplacement depth); geologic and hydrologic description of the site and the conceptual hydrologic model for the site; description of release scenario; hydrologic model implementation and results; engineered barriers and leach rate modeling; transport model implementation and results; and dose model implementation and results. These studies indicate the following: numerical modeling is a practical approach to post-closure safety assessment analysis for nuclear waste repositories; near-field modeling capability needs improvement to permit assessment of the consequences of human intrusion and pumping well scenarios; engineered barrier systems can be useful in mitigating consequences for postulated release scenarios that short-circuit the geohydrologic system; geohydrologic systems separating a repository from the natural biosphere discharge sites act to mitigate the consequences of postulated breaches in containment; and engineered barriers of types other than the containment or absorptive type may be useful.

  4. Solid waste projection model: Model version 1. 0 technical reference manual

    SciTech Connect (OSTI)

    Wilkins, M.L.; Crow, V.L.; Buska, D.E. (Pacific Northwest Lab., Richland, WA (USA)); Ouderkirk, S.J. (Boeing Computer Services Co., Richland, WA (USA))

    1990-11-01T23:59:59.000Z

    The Solid Waste Projection Model (SWPM) system is an analytical tool developed by Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company (WHC). The SWPM system provides a modeling and analysis environment that supports decisions in the process of evaluating various solid waste management alternatives. This document, one of a series describing the SWPM system, contains detailed information regarding the software utilized in developing Version 1.0 of the modeling unit of SWPM. This document is intended for use by experienced software engineers and supports programming, code maintenance, and model enhancement. Those interested in using SWPM should refer to the SWPM Model User's Guide. This document is available from either the PNL project manager (D. L. Stiles, 509-376-4154) or the WHC program monitor (B. C. Anderson, 509-373-2796). 8 figs.

  5. Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro`s Bruce Nuclear Generating Station `A`

    SciTech Connect (OSTI)

    Day, J.E.; Baker, R.L.

    1995-05-01T23:59:59.000Z

    Ontario Hydro at the Bruce Nuclear Generating Station `A` has undertaken a program to render the station`s liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology.

  6. Inter-relation between technical and jurisdictional aspects of hazardous waste management in Houston

    E-Print Network [OSTI]

    Vasavada, Nishith Maheshbhai

    1987-01-01T23:59:59.000Z

    in Houston. (May 1987) Nishith Maheshbhai Vasavada, B. E. , M. S. University M. E. & M. S. University Chair of Advisory Committee: Dr. Roy Harm Problems created by hazardous waste have occurred very fast and have led to promulgation of many regulations... 3. 4, 2 Texas Water Commission 3. 4. 3 Department of Health, City of Houston 3. 4. 4 Community Awareness and Emergency Response (CARR) 3. 4. 5 Houston Chamber of Commerce Recommendations 48 48 48 49 50 51 51 CHAPTER IV ? MUNICIPAL...

  7. Hanford Waste Vitrification Plant technical background document for toxics best available control technology demonstration

    SciTech Connect (OSTI)

    none,

    1992-10-01T23:59:59.000Z

    This document provides information on toxic air pollutant emissions to support the Notice of Construction for the proposed Hanford Waste Vitrification Plant (HWVP) to be built at the the Department of Energy Hanford Site near Richland, Washington. Because approval must be received prior to initiating construction of the facility, state and federal Clean Air Act Notices of construction are being prepared along with necessary support documentation.

  8. Management of Low-Level Radioactive Waste from Research, Hospitals and Nuclear Medical Centers in Egypt - 13469

    SciTech Connect (OSTI)

    Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F. [Hot Labs and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)] [Hot Labs and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)

    2013-07-01T23:59:59.000Z

    The application of radioisotopes and radiation sources in medical diagnosis and therapy is an important issue. Physicians can use radioisotopes to diagnose and treat diseases. Methods of treatment, conditioning and management of low level radioactive wastes from the use of radiation sources and radioisotopes in hospitals and nuclear medicine application, are described. Solid Radioactive waste with low-level activity after accumulation, minimization, segregation and measurement, are burned or compressed in a compactor according to the international standards. Conditioned drums are transported to the interim storage site at the Egyptian Atomic Energy Authority (EAEA) represented in Hot Labs and Waste Management Center (HLWMC) for storage and monitoring. (authors)

  9. NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

    SciTech Connect (OSTI)

    none,

    1982-07-01T23:59:59.000Z

    This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

  10. Corrosion-induced gas generation in a nuclear waste repository: Reactive geochemistry and multiphase flow effect

    SciTech Connect (OSTI)

    Xu, T.; Senger, R.; Finsterle, S.

    2008-10-15T23:59:59.000Z

    Corrosion of steel canisters, stored in a repository for spent fuel and high-level nuclear wastes, leads to the generation and accumulation of hydrogen gas in the backfilled emplacement tunnels, which may significantly affect long-term repository safety. Previous studies used H{sub 2} generation rates based on the volume of the waste or canister material and the stoichiometry of the corrosion reaction. However, iron corrosion and H{sub 2} generation rates vary with time, depending on factors such as amount of iron, water availability, water contact area, and aqueous and solid chemistry. To account for these factors and feedback mechanisms, we developed a chemistry model related to iron corrosion, coupled with two-phase (liquid and gas) flow phenomena that are driven by gas-pressure buildup associated with H{sub 2} generation and water consumption. Results indicate that by dynamically calculating H{sub 2} generation rates based on a simple model of corrosion chemistry, and by coupling this corrosion reaction with two-phase flow processes, the degree and extent of gas pressure buildup could be much smaller compared to a model that neglects the coupling between flow and reactive transport mechanisms. By considering the feedback of corrosion chemistry, the gas pressure increases initially at the canister, but later decreases and eventually returns to a stabilized pressure that is slightly higher than the background pressure. The current study focuses on corrosion under anaerobic conditions for which the coupled hydrogeochemical model was used to examine the role of selected physical parameters on the H{sub 2} gas generation and corresponding pressure buildup in a nuclear waste repository. The developed model can be applied to evaluate the effect of water and mineral chemistry of the buffer and host rock on the corrosion reaction for future site-specific studies.

  11. Social impacts of hazardous and nuclear facilities and events: Implications for Nevada and the Yucca Mountain high-level nuclear waste repository; [Final report

    SciTech Connect (OSTI)

    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)

    1992-05-01T23:59:59.000Z

    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.

  12. Site-specific EIS ordered but injunctive relief deined in nuclear waste storage case

    SciTech Connect (OSTI)

    Barnhart y Chavez, S.

    1980-01-01T23:59:59.000Z

    The Energy Research and Development Administration (ERDA) received appropriations in 1976-77 to construct 22 tanks for storage of high level radioactive wastes generated by its nuclear weapons materials production program. The tanks were to replace older, leaking tanks at the Hanford Reservation in Richland, Washington and the Savannah River Plant in Aiken, South Carolina. The Natural Resources Defense Council (NRDC) had unsuccessfully requested that ERDA obtain a construction permit from the Nuclear Regulatory Commission (NRC). NRDC also petitioned NRC to exercise its licensing authority over the tanks under Section 202(4) of the Energy Reorganization Act of 1974. In response to the NRDC request, ERDA claimed the tanks were only for short-term storage and therefore a license was unnecessary. NRC claimed it lacked jurisdiction over the tanks. NRDC filed suit in United States District Court for the District of Columbia, alleging that ERDA had violated Section 102(2)(C) of the National Environmental Policy Act, and that both ERDA and NRC had violated Section 202(4) of the Energy Reorganization Act. NRDC requested an injunction against further construction of the tanks. Although ERDA did not have to obtain an NRC construction permit for the nuclear waste storage tanks at Hanford Reservation and Savannah River Plant, the programmatic Environmental Impact Statement submitted was insufficient and site-specific statements must be prepared. Injunctive relief pending the statements was denied for the social and economic costs of delaying the tanks project. NRC decisions even remotely connected to its licensing power should be contested in federal courts of appeals, not district courts. The court gave NRDC a hollow victory by ordering a more specific EIS, but denying an injunction.

  13. Hanford Waste Vitrification Plant technical background document for best available radionuclide control technology demonstration

    SciTech Connect (OSTI)

    Carpenter, A.B.; Skone, S.S.; Rodenhizer, D.G.; Marusich, M.V. (Ebasco Services, Inc., Bellevue, WA (USA))

    1990-10-01T23:59:59.000Z

    This report provides the background documentation to support applications for approval to construct and operate new radionuclide emission sources at the Hanford Waste Vitrification Plant (HWVP) near Richland, Washington. The HWVP is required to obtain permits under federal and state statutes for atmospheric discharges of radionuclides. Since these permits must be issued prior to construction of the facility, draft permit applications are being prepared, as well as documentation to support these permits. This report addresses the applicable requirements and demonstrates that the preferred design meets energy, environmental, and economic criteria for Best Available Radionuclide Control Technology (BARCT) at HWVP. 22 refs., 11 figs., 25 tabs.

  14. Coolside waste management research. Quarterly technical progress report, October 1, 1991--October 31, 1991

    SciTech Connect (OSTI)

    Not Available

    1991-12-31T23:59:59.000Z

    Objective was to produce sufficient information on physical and chemical nature of Coolside waste (Coolside No.1, 3 at Edgewater power plant) to design and construct stable, environmentally safe landfills. Progress during this period was centered on analytical method development, elemental and mineralogical analysis of samples, and field facilities preparation to receive lysimeter fill. Sample preparation techniques for thick target PIXE/PIGE were investigated; good agreement between measured and actual values for standard fly ash were obtained for all elements except Fe, Ba, K (PIXE).

  15. Establishing the Technical Basis for Disposal of Heat-generating Waste in

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010Salt | Department of Energy Establishing the Technical

  16. NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547

    SciTech Connect (OSTI)

    Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H. [United States Nuclear Regulatory Commission, Rockville, Maryland 20852 (United States); Pan, Y. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX 78238 (United States)

    2012-07-01T23:59:59.000Z

    Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

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

    SciTech Connect (OSTI)

    Rechard, R.P. [ed.

    1993-12-01T23:59:59.000Z

    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.

  18. Coolside waste management research. Quarterly technical report, April 1--June 30, 1995

    SciTech Connect (OSTI)

    NONE

    1995-12-31T23:59:59.000Z

    This report consists of 3 monthly progress reports. The first represents a summary of results from mineralogical studies of the field lysimeter samples. This part of the project is an ongoing task to understand the long term mineralogical reactions that occur in the lysimeters as a function of static loading (compaction) and moisture content. The data is congruent with results obtained from geotechnical characterization of pre-aged and non-aged Coolside samples with and without surcharge. The investigations are expected to aid in the understanding of the processes that control permeability and leaching potential of the materials and to produce sufficient information on the physical and chemical nature of Coolside waste to design and construct physically stable and environmentally safe landfills. The capacity of various FGD wastes to absorb CO{sub 2} has been recently investigated with the results summarized in the second monthly. The potential usage is for the removal of CO{sub 2} from multi-component gas streams, in particular, natural-gas streams. The third comprises results from ongoing geotechnical testing. The results are concurrent with mineralogical findings that suggest that ettringite, gypsum and calcium-alumino-silicate hydrate phases proceed to form within the aging materials. In specimens with higher degrees of static loading, minerals are forced to grow within available pore space and fractures, which causes less swell. This report also summarizes results from a study of the effects of Coolside leachate on natural clay liners.

  19. Technical assessment of processes to enable recycling of low-level contaminated metal waste

    SciTech Connect (OSTI)

    Reimann, G.A.

    1991-10-01T23:59:59.000Z

    Accumulations of metal waste exhibiting low levels of radioactivity (LLCMW) have become a national burden, both financially and environmentally. Much of this metal could be considered as a resource. The Department of Energy was assigned the task of inventorying and classifying LLCMW, identifying potential applications, and applying and/or developing the technology necessary to enable recycling. One application for recycled LLCMW is high-quality canisters for permanent repository storage of high-level waste (HLW). As many as 80,000 canisters will be needed by 2035. Much of the technology needed to decontaminate LLCMW has already been developed, but no integrated process has been described, even on a pilot scale, for recycling LLCMW into HLW canisters. This report reviews practices for removal of radionuclides and for producing low carbon stainless steel. Contaminants that readily form oxides may be reduced to below de minimis levels and combined with a slag. Most of the radioactivity remaining in the ingot is concentrated in the inclusions. Radionuclides that chemically resemble the elements that comprise stainless steel can not be removed effectively. Slag compositions, current melting practices, and canister fabrication techniques were reviewed.

  20. Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

    SciTech Connect (OSTI)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

    2010-09-01T23:59:59.000Z

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.