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Sample records for accident prevention provisions

  1. DOE Accident Prevention and Investigation Program | Department...

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

    DOE Accident Prevention and Investigation Program DOE Accident Prevention and Investigation Program The Department of Energy (DOE) Accident Prevention and Investigation Program...

  2. DOE Accident Prevention and Investigation Program | Department of Energy

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

    DOE Accident Prevention and Investigation Program DOE Accident Prevention and Investigation Program The Department of Energy (DOE) Accident Prevention and Investigation Program serves as a key DOE corporate safety resource for promoting accident PREVENTION through exchange of lessons learned and information for improvement of our integrated safety management system. The techniques and tools utilized in the investigation of "accidents" can be valuable in looking at leading indicators

  3. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    SciTech Connect (OSTI)

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  4. Role of Passive Safety Systems in Severe Accidents Prevention for Advanced WWER-1000 Reactor Plants

    SciTech Connect (OSTI)

    Bukin, N.V.; Fil, N.S.; Shumsky, A.M. [EDO 'Gidropress', 21 Ordzhonikidze str., Podolsk, Moscow Region, RU-142103 (Russian Federation)

    2004-07-01

    Role of new safety systems applied in advanced WWER-1000 (passive residual heat removal system, SPOT and passive core flooding system, HA-2) in severe accident prevention is considered in the paper. The following typical beyond-design accidents (BDBAs) that essentially determine the design basis of the above passive systems are considered in the paper: - station blackout; - LB LOCA (double-ended cold leg break 850 mm diameter) with station blackout. The domestic DINAMIKA-97 and TETCH-M-97 codes developed by EDO 'Gidropress' were used for the analyses. Besides, some supporting calculations have been performed by new Russian KORSAR code and western RELAP5/MOD3.2 and ATHLET 1.2A codes. The analysis of station blackout accident without operation of new passive systems have shown the exceeding of the maximum design limit of fuel rod damage already in 2-2,5 h after initiating event. Operation of SPOT system prevents any core damage during the BDBA under consideration. The analysis have also demonstrated that operation of new passive safety systems (SPOT and HA-2) ensures the effective core cooling within required period of time. This ensures essentially decreased probability of severe core degradation. (authors)

  5. Accident prevention and Clean Air Act Amendments of 1990 with particular reference to anhydrous hydrogen fluoride

    SciTech Connect (OSTI)

    Kaiser, G.D. (Science Applications International Corp., McLean, VA (United States))

    1993-07-01

    The sections of the Clean Air Act Amendments (CAAA) of 1990 that refer to accident prevention are to be found in Title III. Two significant requirements of the CAAA in this respect relate to the responsibilities of the Occupational Safety and Health Administration (OSHA), which has promulgated a new Process Safety Management (PSM) standard and the Environmental Protection Agency (EPA), which at the time of writing, is developing Risk Management Program (RMP) regulations. The focus of this paper is on how the requirements of the CAAA may affect the reasons for performing a Quantitative Risk Assessment (QRA) or may affect the results of QRA. In order to limit the discussion, this paper focuses on HF. First, the CAAA requires that the EPA assess the hazards associated with HF; the EPA's current draft report is discussed. Second, a generic assessment of the risks associated with the use of HF is given, with emphasis on alkylation units in refineries. The principal contributors to risk are listed. Finally, an assessment of OSHA's PSM standard 29 CFR 1910.119, the related requirements of state laws such as California's Risk Management and Prevention Program and the potential requirement of EPA's Risk Management Program are given, including an assessment of how these requirements may influence quantitative estimates of risk. 13 refs., 1 fig.

  6. Accident Investigations

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

    1996-04-26

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment, safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 2, 4-26-96

  7. Accident Investigations

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

    1995-10-26

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment , safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 1, 10-26-95. Cancels parts of DOE 5484.1

  8. Accident Investigation Handbook

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

    ... the contractor's organization and prevent accidents by ... DOE and Contractor management systems, organizational ... of the Environmental Health and Safety (ES&H) Working Group. ...

  9. Accident management information needs

    SciTech Connect (OSTI)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  10. Federally Led Accident Investigation Reports | Department of...

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

    The U.S. Department of Energy (DOE) Accident Prevention Investigation Board was appointed ... of the June 1, 2013, Stairway Fall Resulting in a Federal Employee Fatality at ...

  11. Accident Investigations

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

    2011-03-04

    This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities. Supersedes DOE O 225.1A. Cancels DOE G 225.1A-1.

  12. Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

    SciTech Connect (OSTI)

    Tusheva, P.; Schaefer, F.; Kliem, S.

    2012-07-01

    The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

  13. Severe Accident Modeling

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

    Severe Accident Modeling - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power ...

  14. Intellectual Property Provisions

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (CLB-115) Cooperative Agreement Research, Development, or Demonstration Large Business and Foreign Entity 01. FAR 52.227-1 Authorization and Consent (DEC 2007) Alternate I (APR 1984) 02. FAR 52.227-2 Notice and Assistance Regarding Patent and Copyright Infringement (DEC 2007) 03. 2 CFR 910 Appendix A of Subpart D Rights in Data - General 04. 2 CFR 910 Appendix A of Subpart D Patent Rights (Large Business Firms - No Waiver) NOTE: In reading these provisions,

  15. Microsoft Word - Unrelated Accident

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

    For Immediate Release Truck Accident Did Not Involve WIPP Shipment CARLSBAD, N.M., October 1, 2009 - A Wednesday night truck accident north of Albuquerque on Highway 165 that involved an 18-wheeler is not related to Waste Isolation Pilot Plant (WIPP) transuranic waste shipments. Involved in the accident was a load of new, unused 55-gallon drums manufactured in Carlsbad that was en route to Richland, Washington. The Waste Isolation Pilot Plant is a U.S. Department of Energy facility designed to

  16. Accident resistant transport container

    DOE Patents [OSTI]

    Andersen, John A. (Albuquerque, NM); Cole, James K. (Albuquerque, NM)

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  17. Accident Response Group | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Accident Response Group | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at

  18. Chernobyl Nuclear Accident | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Chernobyl Nuclear Accident | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working

  19. Attachment C: Optional AIP Provisions

    Broader source: Energy.gov [DOE]

    This attachment contains sample language for a number of optional provisions that sites may choose to include or not to include in their AIPs. Optional AIP provisions are not mandatory and the...

  20. Small Business Solicitation Provision

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

    Solicitation Provision The Department of Energy has many exceptionally complex projects and programs involving the environmental remediation of its sites, and because of this complexity, are beyond the capabilities of small business in terms of a small business acting as the prime contractor. However, there are many opportunities within these projects in which small businesses can take part. It is the Department of Energy's policy to promote the participation of small business in all of its

  1. Intellectual Property Provisions

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (CDLB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity 01. FAR 52.227-1 Authorization and Consent (DEC 2007) Alternate I (APR 1984) 02. FAR 52.227-2 Notice and Assistance Regarding Patent and Copyright Infringement (DEC 2007) 03. 2 CFR 910 Appendix A of Subpart D Rights in Data - Programs Covered under Special Data Statutes 04. 2 CFR 910 Appendix A of Subpart D Patent Rights (Large

  2. Intellectual Property Provisions

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (GDLB-115) Grant - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity 01. 2 CFR 910 Appendix A of Subpart D Rights in Data - Programs Covered under Special Data Statutes 02. 2 CFR 910 Appendix A of Subpart D Patent Rights (Large Business Firms - No Waiver) GDLB-115 1 01. 2 CFR 910, Appendix A of Subpart D, Rights in Data - Programs Covered Under Special Data Statutes (a) Definitions Computer Data Bases, as used in

  3. Intellectual Property Provisions

    Office of Environmental Management (EM)

    GNP-115 1 Intellectual Property Provisions (GNP-115) Grant and Cooperative Agreement Research, Development, or Demonstration Non-Federal Entity (State, Local government, Indian tribe, Institution of higher education, or Nonprofit organization) A Non-Federal Entity is subject to the intellectual property requirements at 2 CFR 200.315. 2 CFR 200.315 Intangible Property (a) Title to intangible property (see §200.59 Intangible property) acquired under a Federal award vests upon acquisition in the

  4. Intellectual Property Provisions - Assistance

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (CDSB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Domestic Small Business 01. FAR 52.227-1 Authorization and Consent (DEC 2007) Alternate I (APR 1984) 02. FAR 52.227-2 Notice and Assistance Regarding Patent and Copyright Infringement (DEC 2007) 03. 2 CFR 910 Appendix A of Subpart D Rights in Data - Programs Covered under Special Data Statutes 04. 2 CFR 910 Appendix A of Subpart D Patent Rights (Small Business Firms

  5. Federally Led Accident Investigation Reports | Department of...

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

    Federally Led Accident Investigation Reports Federally Led Accident Investigation Reports Includes Pre-March 2011 Type A Reports June 1, 1999 Type A Accident Investigation Board...

  6. Intellectual Property Provisions

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (GLB-115) Grant Research, Development, or Demonstration Large Business and Foreign Entity 01. 2 CFR 910 Appendix A of Subpart D Rights in Data - General 02. 2 CFR 910 Appendix A of Subpart D Patent Rights (Large Business Firms - No Waiver) GLB-115 1 01. 2 CFR Part 910, Appendix A of Subpart D, Rights in Data - General (a) Definitions Computer Data Bases, as used in this clause, means a collection of data in a form capable of, and for the purpose of, being

  7. Intellectual Property Provisions

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (GSB-115) Grant Research, Development, or Demonstration Domestic Small Business 01. 2 CFR 910 Appendix A of Subpart D Rights in Data - General 02. 2 CFR 910 Appendix A of Subpart D Patent Rights (Small Business Firms and Nonprofit Organizations) GSB-115 1 01. 2 CFR Part 910, Appendix A of Subpart D, Rights in Data - General (a) Definitions Computer Data Bases, as used in this clause, means a collection of data in a form capable of, and for the purpose of,

  8. Severe Accident Modeling

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

    Severe Accident Modeling - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  9. Severe Accident Studies

    Office of Environmental Management (EM)

    Severe Accident Studies Christopher S. Bajwa Division of Spent Fuel Storage and Transportation Office of Nuclear Material Safety and Safeguards USNRC 2012 U.S. DOE National Transportation Stakeholders Forum (NTSF) May 15 - 17, 2012 Knoxville, TN * Going The Distance? - The Safe Transport of Spent Nuclear Fuel and High-Level Radioactive Waste in the United States * Released February 9, 2006 * Conclusions: * NRC safety regulations are adequate to ensure package containment effectiveness over a

  10. Accident motivates scholarship recipient

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

    Accident motivates scholarship recipient Leyba encourages students: apply for Los Alamos Employees' Scholarship Fund Life-changing experience: springboard to a career in exercise, science, and physical therapy. April 3, 2012 Jacob Leyba, recipient of the Los Alamos Employees' Scholarship Fund Domenici scholarship Jacob Leyba, recipient of the Senator Pete Domenici Endowed Scholarship Fund. Contact Kathy Keith Community Relations & Partnerships (505) 665-4400 Email "I've been through

  11. Accident Investigation of the June 17, 2012, Construction Accident -

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

    Structural Steel Collapse at The Over pack Storage Expansion #2 at the Naval Reactors Facility at the Idaho National Laboratory, Idaho Falls, Idaho | Department of Energy 7, 2012, Construction Accident - Structural Steel Collapse at The Over pack Storage Expansion #2 at the Naval Reactors Facility at the Idaho National Laboratory, Idaho Falls, Idaho Accident Investigation of the June 17, 2012, Construction Accident - Structural Steel Collapse at The Over pack Storage Expansion #2 at the

  12. Accident Investigation of the June 17, 2012, Construction Accident...

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

    at the Idaho National Laboratory, Idaho Falls, Idaho Accident Investigation of the June ... at the Idaho National Laboratory, Idaho Falls, Idaho August 24, 2012 This report ...

  13. DOE Challenge Home Recommended Quality Management Provisions

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

    Recommended Quality Management Provisions RECOMMENDED QUALITY MANAGEMENT PROVISIONS QM Provision Builder Documentation & Verification Requirements Rater/Verifier Requirements QM #1: Project Documentation Complete construction documents shall be qualified as Designed to Earn ENERGY STAR and document all additional provisions and specifications required for DOE Challenge Home including mandatory provisions: 2012 IECC envelope insulation levels, ENERGY STAR windows, duct work in conditioned

  14. Structural assessment of accident loads

    SciTech Connect (OSTI)

    Wagenblast, G.R., Westinghouse Hanford

    1996-05-28

    Structural assessments were made for specific accident loads for specific catch, receiver, and storage tanks. The evaluation herein represents level-of-effort order-of-magnitude estimates of limiting loads that would lead to collapse or rupture of the tank and unmitigated loss of confinement for the waste. Structural capacities were established using failure criteria. Compliance with codes such as ACI, ASCE, ASME, RCRA, UBC, WAC, and DOE Orders was `NOT` maintained. Normal code practice is to prevent failure with margins consistent with expected variations in loads and strengths and confidence in analysis techniques. The evaluation herein represent estimates of code limits without code load factors or code strength reduction factors, and loading beyond such a limit is considered as an onset of some failure mode. The exact nature of the failure mode and its relation to a safe condition is a judgment of the analyst. Consequently, these `RESULTS SHALL NOT BE USED TO ESTABLISH OPERATING OR SAFETY LOAD LIMITS FOR THESE TANKS`.

  15. Guidance for Radiation Accident Management

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

    Procedure Demonstration Introduction Radioactive materials are among the many kinds of hazardous substances emergency responders might have to deal with in an accident. It is prudent that they know their role in responding to a radiation accident should one occur in their communities. The information provided here addresses not only basic explanations and definitions related to radiation but also offers guidance to those responding both at the scene of an accident (prehospital) and at the

  16. Pollution Prevention

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

    Prevention Pollution Prevention Promoting green purchasing, reuse and recycling, and the conservation of fuel, energy, and water. April 17, 2012 Pollution prevention and control...

  17. Intellectual Property Provisions (GDSB-1003)

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

    Attachment #2 GDSB-1003 Intellectual Property Provisions (GDSB-1003) Grant - Special Data Statute Research, Development, or Demonstration Domestic Small Businesses 01. 10 CFR 600.325 Rights in Data - Programs Covered Under Special Data Appendix A Statues (OCT 2003) 02 10 CFR 600.325 Patent Rights (Small Business Firms and Nonprofit Appendix A Organizations) (OCT 2003) 03. FAR 52.227-23 Rights to Proposal Data (Technical) (JUN 1987) NOTE: In reading these provisions, any reference to

  18. First Responders and Criticality Accidents

    SciTech Connect (OSTI)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  19. Severe Accident Studies | Department of Energy

    Energy Savers [EERE]

    Severe Accident Studies Severe Accident Studies Powerpoint discussing studies and conclusions on transportation accidents and safety. PDF icon Severe Accident Studies More Documents & Publications Spent Fuel Transportation Risk Assessment DOE-STD-101-92 EIS-0218-SA-07: Supplement Analysis

  20. Accident tolerant fuel analysis

    SciTech Connect (OSTI)

    Smith, Curtis; Chichester, Heather; Johns, Jesse; Teague, Melissa; Tonks, Michael Idaho National Laboratory; Youngblood, Robert

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  1. Accident Tolerant Fuel Analysis

    SciTech Connect (OSTI)

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  2. Intellectual Property Provisions (CSB-115) Cooperative Agreement...

    Energy Savers [EERE]

    15) Cooperative Agreement Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (CSB-115) Cooperative Agreement Research, Development, or...

  3. Intellectual Property Provisions (CLB-115) Cooperative Agreement...

    Office of Environmental Management (EM)

    Agreement Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (CLB-115) Cooperative Agreement Research, Development, or...

  4. Environment/Health/Safety (EHS): Monthly Accident Statistics

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

    Personal Protective Equipment (PPE) Injury Review & Analysis Worker Safety and Health Program: PUB-3851 Monthly Accident Statistics Latest Accident Statistics Accident...

  5. Pollution Prevention

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

    Pollution Prevention Goal 5: Pollution Prevention LANL is dedicated to finding ways to reduce waste, prevent pollution, and recycle waste that cannot be reduced. Energy Conservation» Efficient Water Use & Management» High Performance Sustainable Buildings» Greening Transportation» Green Purchasing & Green Technology» Pollution Prevention» Science Serving Sustainability» ENVIRONMENTAL SUSTAINABILITY GOALS at LANL Technical Area 21: Water was sprayed during the demolition of 24 Cold

  6. Pollution Prevention

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

    Pollution Prevention Pollution Prevention Promoting green purchasing, reuse and recycling, and the conservation of fuel, energy, and water. April 17, 2012 Pollution prevention and control at LANL Contact Environmental Communication & Public Involvement P.O. Box 1663 MS M996 Los Alamos, NM 87545 (505) 667-0216 Email Our goal is to reduce or eliminate waste whenever possible. Promoting pollution prevention to achieve sustainability Our commitment to environmental stewardship and sustainability

  7. ORISE: REAC/TS Radiation Accident Registries

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

    Accident Registries The Radiation Emergency Assistance Center/Training Site (REAC/TS) at the Oak Ridge Institute for Science and Education (ORISE) maintains a number of radiation accident registries that provide medical professionals with up-to-date radiation accident information. Information for these accident registries is gathered from many sources, including the World Health Organization, International Atomic Energy Agency, U.S. Nuclear Regulatory Commission, state radiological health

  8. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    SciTech Connect (OSTI)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  9. K Basins floor sludge retrieval system knockout pot basket fuel burn accident

    SciTech Connect (OSTI)

    HUNT, J.W.

    1998-11-11

    The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool.

  10. Small Business Soliciation Provision | Department of Energy

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

    Soliciation Provision Small Business Soliciation Provision The Department of Energy has many exceptionally complex projects and programs involving the environmental remediation of its sites, and because of this complexity, are beyond the capabilities of small business in terms of a small business acting as the prime contractor. However, there are many opportunities within these projects in which small businesses can take part. PDF icon Small Business Solicitation Provision.pdf More Documents

  11. Intellectual Property Provisions (CDLB-115) Cooperative Agreement...

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

    15.pdf More Documents & Publications CDLB-1003.PDF&0; Intellectual Property Provisions (CLB-115) Cooperative Agreement Research, Development, or Demonstration Large Business and...

  12. Intellectual Property Provisions (CDSB-115) Cooperative Agreement...

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

    15.pdf More Documents & Publications CDSB-1003.PDF&0; Intellectual Property Provisions (CDLB-115) Cooperative Agreement - Special Data Statute Research, Development, or...

  13. DOE Challenge Home Recommended Quality Management Provisions...

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

    DOE Challenge Home Recommended Quality Management Provisions. PDF icon qm6-14-13.pdf More Documents & Publications Version Tracking Document for DOE Challenge Homes, National ...

  14. Microsoft Word - Intellectual Property Provisions _NRD-1003_...

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

    More Documents & Publications Intellectual Property Provisions (NRD-115) Nonresearch and Development Microsoft Word - ARRAAttachment12v1.doc D:WebGCCurrentMaterialtechtrans...

  15. Accident analysis and DOE criteria

    SciTech Connect (OSTI)

    Graf, J.M.; Elder, J.C.

    1982-01-01

    In analyzing the radiological consequences of major accidents at DOE facilities one finds that many facilities fall so far below the limits of DOE Order 6430 that compliance is easily demonstrated by simple analysis. For those cases where the amount of radioactive material and the dispersive energy available are enough for accident consequences to approach the limits, the models and assumptions used become critical. In some cases the models themselves are the difference between meeting the criteria or not meeting them. Further, in one case, we found that not only did the selection of models determine compliance but the selection of applicable criteria from different chapters of Order 6430 also made the difference. DOE has recognized the problem of different criteria in different chapters applying to one facility, and has proceeded to make changes for the sake of consistency. We have proposed to outline the specific steps needed in an accident analysis and suggest appropriate models, parameters, and assumptions. As a result we feed DOE siting and design criteria will be more fairly and consistently applied.

  16. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    SciTech Connect (OSTI)

    S.O. Bader

    1999-10-18

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.

  17. Cold Vacuum Drying facility design basis accident analysis documentation

    SciTech Connect (OSTI)

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

  18. HTGR severe accident sequence analysis

    SciTech Connect (OSTI)

    Harrington, R.M.; Ball, S.J.; Kornegay, F.C.

    1982-01-01

    Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.

  19. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-07-12

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  20. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  1. ORISE: REAC/TS Radiation Accident Registries

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

    for these accident registries is gathered from many sources, including the World Health Organization, International Atomic Energy Agency, U.S. Nuclear Regulatory Commission,...

  2. EMERGENCY RESPONSE TO A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL

    Office of Environmental Management (EM)

    Emer Emer Emer Emer Emergency Response to a T gency Response to a T gency Response to a T gency Response to a T gency Response to a Transportation ransportation ransportation ransportation ransportation Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER Viewing this video and completing the

  3. ADR Provisions in Federal Acquisition Regulation (FAR) | Department of

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

    Energy ADR Provisions in Federal Acquisition Regulation (FAR) » ADR Provisions in Federal Acquisition Regulation (FAR) ADR Provisions in Federal Acquisition Regulation (FAR) ADR Provisions in Federal Acquisition Regulation (FAR) PDF icon ADR Provisions in Federal Acquisition Regulation (FAR) More Documents & Publications Microsoft Word - ADR-Provisions-FederalAcquisitionReg.doc ACQUISITION LETTER DEAR Part 933 Microsoft Word - ACQUISITION LETTER.doc

  4. Light-water reactor accident classification

    SciTech Connect (OSTI)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  5. Attachment D: Mandatory Provisions if Applicable

    Broader source: Energy.gov [DOE]

    This attachment contains provisions that may be mandatory for an AIP if the State: (1) chooses to develop an emergency response program but does not want to use the suggested language in Attachment...

  6. Commercial SNF Accident Release Fractions

    SciTech Connect (OSTI)

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the container that confines the fuel assemblies could provide an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. This analysis, however, does not take credit for the additional barrier and establishes only the total release fractions for bare unconfined intact commercial SNF assemblies, which may be conservatively applied to confined intact commercial I SNF assemblies.

  7. Bike-Sharing:History, Impacts, Models of Provision, and Future...

    Open Energy Info (EERE)

    Bike-Sharing:History, Impacts, Models of Provision, and Future Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Bike-Sharing:History, Impacts, Models of Provision, and...

  8. Guidance on Meeting Executive Order 13693 Water Provisions |...

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

    Water Efficiency Guidance on Meeting Executive Order 13693 Water Provisions Guidance on Meeting Executive Order 13693 Water Provisions Executive Order (E.O.) 13693 became ...

  9. Section 1605 Buy American Provision under ARRA | Department of...

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

    Additional information on the Buy American Provision under ARRA. Section 1605 Buy American Provision under ARRA More Documents & Publications Nationwide Limited Public Interest...

  10. Final Rule on Amending Eligibility Provisions to Multifamily...

    Energy Savers [EERE]

    Final Rule on Amending Eligibility Provisions to Multifamily Buildings for the Weatherization Assistance Program Final Rule on Amending Eligibility Provisions to Multifamily...

  11. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect (OSTI)

    Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

    2012-11-01

    This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

  12. ADR Provisions in Federal Acquisition Regulation (FAR)

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

    ADR Provisions in Federal Acquisition Regulation (FAR) d. Alternative means of dispute resolution. Notwithstanding any other provision of this chapter, a contractor and a contracting officer may use any alternative means of dispute resolution under subchapter IV of chapter 5 of title 5, or other mutually agreeable procedures, for resolving claims. In a case in which such alternative means of dispute resolution or other mutually agreeable procedures are used, the contractor shall certify that the

  13. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Savers [EERE]

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to...

  14. Type B Accident Investigation of the July 14, 2005, Americium...

    Office of Environmental Management (EM)

    14, 2005, Americium Contamination Accident at the Sigma Facility, Los Alamos National Laboratory Type B Accident Investigation of the July 14, 2005, Americium Contamination...

  15. Accident Investigation of the February 7, 2013, Scissor Lift...

    Office of Environmental Management (EM)

    February 7, 2013, Scissor Lift Accident in the West Hackberry Brine Tank-14 Resulting in Injury, Strategic Petroleum Reserve West Hackberry, LA Accident Investigation of the...

  16. ORISE: The Medical Basis for Radiation-Accident Preparedness...

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

    The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REACTS Symposium on the Medical Basis for Radiation-Accident...

  17. Release fractions for Rocky Flats specific accidents

    SciTech Connect (OSTI)

    Weiss, R.C.

    1992-09-01

    As Rocky Flats and other DOE facilities begin the transition process towards decommissioning, the nature of the scenarios to be studied in safety analysis will change. Whereas the previous emphasis in safety accidents related to production, now the emphasis is shifting to accidents related tc decommissioning and waste management. Accident scenarios of concern at Rocky Flats now include situations of a different nature and different scale than are represented by most of the existing experimental accident data. This presentation will discuss approaches@to use for applying the existing body of release fraction data to this new emphasis. Mention will also be made of ongoing efforts to produce new data and improve the understanding of physical mechanisms involved.

  18. Crediting Tritium Deposition in Accident Analysis

    SciTech Connect (OSTI)

    Murphy, C.E. Jr.

    2001-06-20

    This paper describes the major aspects of tritium dispersion phenomenology, summarizes deposition attributes of the computer models used in the DOE Complex for tritium dispersion, and recommends an approach to account for deposition in accident analysis.

  19. Recommendations for Analyzing Accidents Under NEPA | Department of Energy

    Energy Savers [EERE]

    Analyzing Accidents Under NEPA Recommendations for Analyzing Accidents Under NEPA This DOE guidance clarifies and supplements "Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements." It focuses on principles of accident analyses under NEPA. PDF icon RECOMMENDATIONS for ANALYZING ACCIDENTS under the NATIONAL ENVIRONMENTAL POLICY ACT More Documents & Publications Recommendations for the Preparation of Environmental Assessments and

  20. RADIOACTIVE WASTE MANAGEMENT IN THE CHERNOBYL EXCLUSION ZONE - 25 YEARS SINCE THE CHERNOBYL NUCLEAR POWER PLANT ACCIDENT

    SciTech Connect (OSTI)

    Farfan, E.; Jannik, T.

    2011-10-01

    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.

  1. SAF-230DE - Web Based Course: Accident Investigation Overview | Department

    Energy Savers [EERE]

    of Energy SAF-230DE - Web Based Course: Accident Investigation Overview SAF-230DE - Web Based Course: Accident Investigation Overview September 18, 2013 - 10:52am Addthis SAF-230DE - Web Based Course: Accident Investigation Overview The Office of Health Safety and Security (HSS) National Training Center (NTC) in collaboration with the HSS Accident Investigation Program (HS-24) has developed and released a course that provides an overview of the fundamentals of accident investigation. This

  2. Phase II Accident Investigation Board Briefing | Department of Energy

    Energy Savers [EERE]

    Phase II Accident Investigation Board Briefing Phase II Accident Investigation Board Briefing Topic: Ted Wyka DOE, Provided a Brief on the Findings in the WIPP Accident Investigation. Information Provided included the Judgments of NEED and Causes That Contributed to the Incident. PDF icon AIB Brief - May 20, 2015 More Documents & Publications Accident Investigation Report Phase II Accident Investigation Report - Radiological Release WIPP Recovery Progress

  3. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    SciTech Connect (OSTI)

    Sdouz, Gert [ARC Seibersdorf Research GmbH, Viktor Kaplan-Strasse 2, 2700 Wr. Neustadt (Austria)

    2006-07-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a 'Large Break LOCA'-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a 'Station Blackout'-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors. (authors)0.

  4. Service experience, structural integrity, severe accidents, and erosion in nuclear and fossil plants. PVP-Volume 303

    SciTech Connect (OSTI)

    Paterson, S.R.; Bamford, W.H; Geraets, L.H.; Okazaki, M.; Cipolla, R.C.; Cowfer, C.D.; Means, K.H.

    1995-12-01

    The objective of this symposium was to disseminate information on service degradation and its prevention. Papers have been divided into the following topical sections: Service experience in nuclear plants; DOE high-level waste tank structural integrity panel--Summary reports; Severe accidents; Service experience in operating fossil power plants; and Erosion. Papers have been processed separately for inclusion on the data base.

  5. Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) Accident Prevention Investigation Board was appointed to investigate a fire at the Waste Isolation Pilot Plant that occurred on February 5, 2014. An aged EIMCO 985-T15 salt haul truck (dump truck) caught fire in an underground mine.

  6. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect (OSTI)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  7. A Review of Criticality Accidents 2000 Revision

    SciTech Connect (OSTI)

    Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

    2000-05-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

  8. Type B Accident Investigation of the July 31, 2006, Fall from...

    Office of Environmental Management (EM)

    Ladder Accident at the Lawrence Livermore National Laboratory, Livermore, California Type B Accident Investigation of the July 31, 2006, Fall from Ladder Accident at the Lawrence...

  9. Type B Accident Investigation Board Report of the April 23, 1997...

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

    April 23, 1997, Helicopter Accident at Raton Pass, Raton Pass, Colorado Type B Accident Investigation Board Report of the April 23, 1997, Helicopter Accident at Raton Pass, Raton...

  10. Top Six Areawide Contract Provisions Forgotten

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

    SIX AREAWIDE CONTRACT PROVISIONS FORGOTTEN LINDA L. COLLINS Hosted by: FEDERAL UTILITY PARTNERSHIP WORKING GROUP SEMINAR November 5-6, 2014 Cape Canaveral. Florida ONE * Term of Authorization May Extend Beyond Term of Areawide Contract Federal Utility Partnership Working Group November 5-6, 2014 Cape Canaveral, FL TWO * Utility Services Are Not Construction Contracts Federal Utility Partnership Working Group November 5-6, 2014 Cape Canaveral, FL THREE * Areawide Contract Holders to notify GSA of

  11. LESSONS LEARNED FROM A RECENT LASER ACCIDENT

    SciTech Connect (OSTI)

    Woods, Michael; /SLAC

    2011-01-26

    A graduate student received a laser eye injury from a femtosecond Ti:sapphire laser beam while adjusting a polarizing beam splitter optic. The direct causes for the accident included failure to follow safe alignment practices and failure to wear the required laser eyewear protection. Underlying root causes included inadequate on-the-job training and supervision, inadequate adherence to requirements, and inadequate appreciation for dimly visible beams outside the range of 400-700nm. This paper describes how the accident occurred, discusses causes and lessons learned, and describes corrective actions being taken.

  12. Fuel performance during severe accidents. [PWR

    SciTech Connect (OSTI)

    Buescher, B.J.; Gruen, G.E.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. This program is underway in the Power Burst Facility at the Idaho National Engineering Laboratory. In preparation for the first test, predictions have been performed using the TRAC-BD1 computer. This paper presents the calculated results showing a slow heatup to 2400 K over 5 hours, and the analysis includes accelerated oxidation of the zirconium cladding at temperatures above 1850 K.

  13. Intellectual Property Provisions (GDSB-115) Grant - Special Data...

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

    - Special Data Statute Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (GDSB-115) Grant - Special Data Statute Research,...

  14. Standard Intellectual Property (IP) Provisions for Financial Assistance

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

    Awards | Department of Energy Standard Intellectual Property (IP) Provisions for Financial Assistance Awards Standard Intellectual Property (IP) Provisions for Financial Assistance Awards Click on Set Number to Access Provisions: Standard Intellectual Property (IP) Provisions for Financial Assistance Awards issued ON OR AFTER December 26, 2014 Type of Award Type of Project Special Data Statute e.g., EPACT) Type of Recipient Set Number (PDF) Cooperative Agreement Research, Development, or

  15. DOE Challenge Home Recommended Quality Management Provisions | Department

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

    of Energy Recommended Quality Management Provisions DOE Challenge Home Recommended Quality Management Provisions DOE Challenge Home Recommended Quality Management Provisions. PDF icon qm_6-14-13.pdf More Documents & Publications Version Tracking Document for DOE Challenge Homes, National Program Requirements (Rev. 03) Washington DOE ZERH Program Requirements DOE Challenge Home, California

  16. Regulations for Implementing the Procedural Provisions of NEPA | Department

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

    of Energy Regulations for Implementing the Procedural Provisions of NEPA Regulations for Implementing the Procedural Provisions of NEPA PDF icon CEQ Regulations for Implementing the Procedural Provisions of NEPA (40 CFR 1500-1508) More Documents & Publications Effective Public Participation Forty Most Asked Questions Concerning CEQ's National Environmental Policy Act Regulations A Citizen's Guide to the NEPA: Having Your Voice Heard

  17. Data Provision Instructions for All DOE Geothermal Technologies Office

    Office of Environmental Management (EM)

    Funds Recipients | Department of Energy Data Provision Instructions for All DOE Geothermal Technologies Office Funds Recipients Data Provision Instructions for All DOE Geothermal Technologies Office Funds Recipients DATA PROVISION INSTRUCTIONS FOR ALL DOE GEOTHERMAL TECHNOLOGIES OFFICE FUNDS RECIPIENTS As required by DOE Geothermal Technologies Office (GTO) Funding Opportunity Announcements for Financial Assistance awards, all GTO funds recipients including national laboratories must provide

  18. PNNL Results from 2009 Silene Criticality Accident Dosimeter Intercomparison Exercise

    SciTech Connect (OSTI)

    Hill, Robin L.; Conrady, Matthew M.

    2010-06-30

    This document reports the results of testing of the Hanford Personnel Nuclear Accident Dosimeter (PNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on October 13, 14, and 15, 2009.

  19. Y-12's 1958 nuclear criticality accident and increased safety...

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

    accident and increased safety - 1958 brought accidents, more safety The first X-ray machine was brought to Y-12 in February, 1949. It was a 1,000 KV system installed in Building...

  20. Severe Accident Test Station Activity Report

    SciTech Connect (OSTI)

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000șC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  1. Accident Investigation Reports - Type B | Department of Energy

    Energy Savers [EERE]

    Accident Investigation Reports - Type B Accident Investigation Reports - Type B November 23, 2010 Type B Accident Investigation Board Report of the September 29, 2010, Radiological Contamination Event at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna, New, York This report is an independent product of the Type B Accident Investigation Board appointed by Mark A. Gilbertson, Deputy Assistant Secretary for Program and Site Support, U.S. Department of Energy.

  2. Computerized Accident Incident Reporting System | Department of Energy

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

    Computerized Accident Incident Reporting System Computerized Accident Incident Reporting System CAIRS Database The Computerized Accident/Incident Reporting System is a database used to collect and analyze DOE and DOE contractor reports of injuries, illnesses, and other accidents that occur during DOE operations. CAIRS is a Government computer system and, as such, has security requirements that must be followed. Access to the database is open to DOE and DOE contractors. Additional information

  3. Development of Light Water Reactor Fuels with Enhanced Accident Tolerance

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

    - Report to Congress | Department of Energy Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress This report provides DOE's plan to develop light water reactor (LWR) fuels with enhanced accident tolerance in response to 2012 Congressional direction and funding authorization. The result of the accident tolerant fuel development activities, if successful,

  4. Naval Spent Fuel Rail Shipment Accident Exercise Objectives | Department of

    Office of Environmental Management (EM)

    Energy Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives PDF icon Naval Spent Fuel Rail Shipment Accident Exercise Objectives More Documents & Publications TEC Meeting Summaries - April 2005 Presentations TEC Meeting Summaries - January - February 2007 Presentations NTSF 2014 Meeting Agenda

  5. Calculation notes that support accident scenario and consequence development for the steam intrusion from interfacing systems accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-07-25

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Steam Intrusion from Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included within.

  6. REAC/TS Radiation Accident Registry: An Overview

    SciTech Connect (OSTI)

    Doran M. Christensen, DO, REAC /TS Associate Director and Staff Physician Becky Murdock, REAC/TS Registry and Health Physics Technician

    2012-12-12

    Over the past four years, REAC/TS has presented a number of case reports from its Radiation Accident Registry. Victims of radiological or nuclear incidents must meet certain dose criteria for an incident to be categorized as an “accident” and be included in the registry. Although the greatest numbers of “accidents” in the United States that have been entered into the registry involve radiation devices, the greater percentage of serious accidents have involved sealed sources of one kind or another. But if one looks at the kinds of accident scenarios that have resulted in extreme consequence, i.e., death, the greater share of deaths has occurred in medical settings.

  7. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

    2014-10-01

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.

  8. US Department of Energy Chernobyl accident bibliography

    SciTech Connect (OSTI)

    Kennedy, R A; Mahaffey, J A; Carr, F Jr

    1992-04-01

    This bibliography has been prepared by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) Office of Health and Environmental Research to provide bibliographic information in a usable format for research studies relating to the Chernobyl nuclear accident that occurred in the Ukrainian Republic, USSR in 1986. This report is a product of the Chernobyl Database Management project. The purpose of this project is to produce and maintain an information system that is the official United States repository for information related to the accident. Two related products prepared for this project are the Chernobyl Bibliographic Search System (ChernoLit{trademark}) and the Chernobyl Radiological Measurements Information System (ChernoDat). This report supersedes the original release of Chernobyl Bibliography (Carr and Mahaffey, 1989). The original report included about 2200 references. Over 4500 references and an index of authors and editors are included in this report.

  9. Risk Estimation Methodology for Launch Accidents.

    SciTech Connect (OSTI)

    Clayton, Daniel James; Lipinski, Ronald J.; Bechtel, Ryan D.

    2014-02-01

    As compact and light weight power sources with reliable, long lives, Radioisotope Power Systems (RPSs) have made space missions to explore the solar system possible. Due to the hazardous material that can be released during a launch accident, the potential health risk of an accident must be quantified, so that appropriate launch approval decisions can be made. One part of the risk estimation involves modeling the response of the RPS to potential accident environments. Due to the complexity of modeling the full RPS response deterministically on dynamic variables, the evaluation is performed in a stochastic manner with a Monte Carlo simulation. The potential consequences can be determined by modeling the transport of the hazardous material in the environment and in human biological pathways. The consequence analysis results are summed and weighted by appropriate likelihood values to give a collection of probabilistic results for the estimation of the potential health risk. This information is used to guide RPS designs, spacecraft designs, mission architecture, or launch procedures to potentially reduce the risk, as well as to inform decision makers of the potential health risks resulting from the use of RPSs for space missions.

  10. Hanford waste tank bump accident analysis

    SciTech Connect (OSTI)

    MALINOVIC, B.

    2003-03-21

    This report provides a new evaluation of the Hanford tank bump accident analysis (HNF-SD-Wh4-SAR-067 2001). The purpose of the new evaluation is to consider new information and to support new recommendations for final safety controls. This evaluation considers historical data, industrial failure modes, plausible accident scenarios, and system responses. A tank bump is a postulated event in which gases, consisting mostly of water vapor, are suddenly emitted from the waste and cause tank headspace pressurization. A tank bump is distinguished from a gas release event in two respects: First, the physical mechanism for release involves vaporization of locally superheated liquid, and second, gases emitted to the head space are not flammable. For this reason, a tank bump is often called a steam bump. In this report, even though non-condensible gases may be considered in bump models, flammability and combustion of emitted gases are not. The analysis scope is safe storage of waste in its current configuration in single-shell tanks (SSTs) and double-shell tanks (DSTs). The analysis considers physical mechanisms for tank bump to formulate criteria for bump potential, application of the criteria to the tanks, and accident analysis of bump scenarios. The result of consequence analysis is the mass of waste released from tanks for specific scenarios where bumps are credible; conversion to health consequences is performed elsewhere using standard Hanford methods (Cowley et al. 2000). The analysis forms a baseline for future extension to consider waste retrieval.

  11. DOE Zero Energy Ready Home Recommended Quality Management Provisions |

    Energy Savers [EERE]

    Department of Energy Recommended Quality Management Provisions DOE Zero Energy Ready Home Recommended Quality Management Provisions DOE Zero Energy Ready Home Recommended Quality Management Provisions. PDF icon QM Cheklist.pdf More Documents & Publications Version Tracking Document for DOE Challenge Homes, National Program Requirements (Rev. 03) Washington DOE ZERH Program Requirements DOE Zero Energy Ready Home Solar Hot Water-Ready Checklist

  12. Bodman Statement On Senate Approval of ANWR Provisions | Department of

    Energy Savers [EERE]

    Energy On Senate Approval of ANWR Provisions Bodman Statement On Senate Approval of ANWR Provisions March 16, 2005 - 10:50am Addthis Washington, DC - Energy Secretary Samuel W. Bodman released the following statement today regarding Senate approval of ANWR provisions in the budget resolution: "The Senate's vote today to clear the way for environmentally responsible oil and gas exploration in a small portion of ANWR is a victory for American consumers, America's economy and America's

  13. UF6 overfilling prevention at Eurodif production Georges Besse plant

    SciTech Connect (OSTI)

    Reneaud, J.M.

    1991-12-31

    Risk of overfilling exists on different equipments of Georges BESSE Plant: cylinders, desublimers and intermediate tanks. The preventive measures are composed of technical devices: desublimers weighing, load monitoring alarms, automatic controls ... and procedures, training, safety organization. In thirteen years of operation, some incidents have occurred but none of them has caused any personal injuries. They are related and discussed. The main factors involved in the Sequoyah fuel facility accident on 1/4/1986 have been analyzed and taken into account.

  14. Intellectual Property Provisions (GLB-115) Grant Research, Development...

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

    GLB-115) Grant Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (GLB-115) Grant Research, Development, or Demonstration...

  15. Intellectual Property Provisions (GDLB-115) Grant - Special Data...

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

    Data Statute Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (GDLB-115) Grant - Special Data Statute Research,...

  16. Intellectual Property Provisions (GSB-115) Grant Research, Development...

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

    GSB-115) Grant Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (GSB-115) Grant Research, Development, or Demonstration Domestic...

  17. Guidelines for Provision and Interchange of Geothermal Data Assets

    Broader source: Energy.gov [DOE]

    This document presents guidelines related to provision and interchange of data assets in the context of the National Geothermal Data System.

  18. DOE Zero Energy Ready Home Recommended Quality Management Provisions...

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

    Management Provisions. PDF icon QM Cheklist.pdf More Documents & Publications Version Tracking Document for DOE Challenge Homes, National Program Requirements (Rev. 03)...

  19. 2 CCR State Lands Commission Article 1, General Provisions |...

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- RegulationRegulation: 2 CCR State Lands Commission Article 1, General ProvisionsLegal Abstract California...

  20. WC 26 - Water Quality Control Administrative Provisions | Open...

    Open Energy Info (EERE)

    26 - Water Quality Control Administrative Provisions Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- RegulationRegulation: WC 26 - Water Quality...

  1. Intellectual Property Provisions (GNP-115) Grant and Cooperative...

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

    Non-Federal Entity (State, Local government, Indian tribe, Institution of higher education, or Nonprofit organization) Intellectual Property Provisions (GNP-115) Grant...

  2. PART FIVE COMMON PROVISIONS ARTICLE XXIX. RECOVERY OF STATE...

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

    COMMON PROVISIONS ARTICLE XXIX. RECOVERY OF STATE COSTS 99. DOE agrees to reimburse Ecology for all of its costs related to the implementation of this Agreement as provided...

  3. WAPA General Power Contract Provisions | Department of Energy

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

    PDF icon WAPAGeneralPowerContractProvisions.pdf More Documents & Publications WAPA Purchase of Energy Contractor Template SPR Pro Forma Contract Audit Report: IG-0409

  4. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    SciTech Connect (OSTI)

    Robb, Kevin R.

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.

  5. Accident Investigation Report Phase II | Department of Energy

    Office of Environmental Management (EM)

    Accident Investigation Report Phase II Accident Investigation Report Phase II On February 14, 2014, an airborne radiological release occurred at the Department of Energy Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. On March 4, 2014, an Accident Investigation Board (the Board) was appointed by Matthew Moury, Deputy Assistant Secretary, Safety, Security, and Quality Programs to determine the cause of the release. Because access to the underground was restricted following the

  6. ORISE: The Medical Basis for Radiation-Accident Preparedness: Medical

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

    Management (Published by REAC/TS) The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REAC/TS Symposium on the Medical Basis for Radiation-Accident Preparedness and the Biodosimetry Workshop As part of its mission to provide continuing education for personnel responsible for treating radiation injuries, REAC/TS hosted the Fifth International REAC/TS Symposium on the Medical Basis for Radiation-Accident Preparedness symposium and

  7. Accident Investigation Report - Fire Report | Department of Energy

    Energy Savers [EERE]

    Fire Report Accident Investigation Report - Fire Report On February 7, 2014, Deputy Assistant Secretary, Safety, Security, and Quality Programs Environmental Management, DOE, formally appointed an Accident Investigation Board to investigate an underground mine fire involving a salt haul truck occurred at DOE's WIPP near Carlsbad, New Mexico. The Board began the investigation on February 10, 2014, and the report is now final and available for the public. PDF icon Accident Investigation Report

  8. Accident Investigations of the February 14, 2014, Radiological Release at

    Energy Savers [EERE]

    the Waste Isolation Pilot Plant, Carlsbad, NM | Department of Energy Accident Investigations of the February 14, 2014, Radiological Release at the Waste Isolation Pilot Plant, Carlsbad, NM Accident Investigations of the February 14, 2014, Radiological Release at the Waste Isolation Pilot Plant, Carlsbad, NM February 14, 2014 Accident Investigations of the February 14, 2014, Radiological Release at the Waste Isolation Pilot Plant, Carlsbad, NM On February 14, 2014, at approximately 2314

  9. Los Alamos National Laboratory Accident Investigation Board Corrective

    Energy Savers [EERE]

    Action Plan Update | Department of Energy Accident Investigation Board Corrective Action Plan Update Los Alamos National Laboratory Accident Investigation Board Corrective Action Plan Update Topic: Status of the Corrective Actions that were identified by the Accident Investigation Board. It was noted that there are 22 Judgments of Need that were assessed against the Los Alamos Site. PDF icon AIB-CAP-Update - January 13, 2016 More Documents & Publications Environmental Management

  10. International Perspective on Fukushima Accident | Department of Energy

    Office of Environmental Management (EM)

    International Perspective on Fukushima Accident International Perspective on Fukushima Accident September 19-20, 2012 Presenter: Miroslav LipĂĄr, Head, Operational Safety Section Department of Nuclear Safety and Security International Atomic Energy Agency Topic Covered: The IAEA before Fukushima -Severe accidents management The IAEA actions after Fukushima The IAEA Action plan on nuclear safety Measures to improve operational safety Conclusions PDF icon International Perspective on Fukushima

  11. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    SciTech Connect (OSTI)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  12. Intellectual Property Provisions (CSB-115) Cooperative Agreement Research,

    Energy Savers [EERE]

    Development, or Demonstration Domestic Small Business | Department of Energy 15) Cooperative Agreement Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (CSB-115) Cooperative Agreement Research, Development, or Demonstration Domestic Small Business PDF icon CSB-115.pdf More Documents & Publications Intellectual Property Provisions (CSB-1003) Cooperative Agreement Research, Development, or Demonstration Domestic Small Businesses Intellectual

  13. Type B Accident Investigation Board Report, May 8, 2004, Exothermic...

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

    (exothermic reaction) accident occurred during heating of surplus activated sodium shields at the East Tennessee Technology Park (ETTP). The work activities were being ...

  14. Accident Investigation of the July 30, 2013, Electrical Fatality...

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

    The scope of the investigation included gathering and documenting all relevant facts of the accident, conducting interviews, review of employee statements, work procedures, ...

  15. Type B Accident Investigation Report on the Exertional Heat Illnesses...

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

    New Mexico, July 13, 2006 Type B Accident Investigation Report on the Exertional Heat Illnesses during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July ...

  16. Type A Accident Investigation of the March 16, 2000, Plutonium...

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

    New Mexico Type A Accident Investigation of the March 16, 2000, Plutonium-238 Multiple Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico July ...

  17. Accident Investigation Report - Fire Report | Department of Energy

    Office of Environmental Management (EM)

    an Accident Investigation Board to investigate an underground mine fire involving a salt haul truck occurred at DOE's WIPP near Carlsbad, New Mexico. The Board began the...

  18. Accident Investigation Reports - Type B | Department of Energy

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

    is an independent product of the Type B Accident Investigation Board appointed by John Kennedy, Acting Manager, Chicago Operations Office, U.S. Department of Energy (DOE). October...

  19. Accident Investigation of the September 20, 2012 Fatal Fall from...

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

    September 20, 2012 Fatal Fall from the Dworshak-Taft 1 Transmission Tower, at the Bonneville Power Marketing Administration Accident Investigation of the September 20, 2012 Fatal ...

  20. Type B Accident Investigation of the Savannah River Site Arc...

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

    Type B Accident Investigation Board investigation of the September 23, 2009, employee burn injury at the Department of Energy (DOE) Savannah River Site (SRS) D Area powerhouse....

  1. Type A Accident Investigation of the June 21, 2001, Drilling...

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

    June 21, 2001, Drilling Rig Operator Injury at the Fermi National Accelerator Laboratory, August 2001 Type A Accident Investigation of the June 21, 2001, Drilling Rig Operator ...

  2. Accident Investigation of the December 11, 2013, Integrated Device...

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

    Pilot Plant, Carlsbad NM Type A Accident Investigation Board Report on the July 11, 1996, Electrical Shock at Technical Area 53, Building MPF-14, Los Alamos National Laboratory

  3. Sandia Assists NASA in Understanding Launch-Area Accidents

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

    Assists NASA in Understanding Launch-Area Accidents - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable ...

  4. Accident Investigation of the August 21, 2012, Contamination...

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

    Security, LLC. The Operating Contractor quickly determined that the contamination had spread offsite, and response teams were immediately brought in. PDF icon Accident...

  5. Type B Accident Investigation Report of the October 28, 2004...

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

    Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA Type B Accident Investigation Report of the October 28, 2004, Burn...

  6. Type B Accident Investigation of the Arc Flash at Brookhaven...

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

    (INL) was investigated in which a technician sustained a serious injury to his right hand while operating a table saw. In conducting its investigation, the Accident...

  7. Type B Accident Investigation Board Report for the January 11...

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

    Idaho Falls, Idaho Type B Accident Investigation Board Report for the January 11, 2006, Personal Injury During Table Saw Use at the Heyrend Way Facility, Idaho Falls, Idaho ...

  8. Type B Accident Investigation Board Report on the October 15...

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

    15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National ... Type B Accident Investigation Board Report on the October 15, 2001, Grout Injection ...

  9. Recommendations for Analyzing Accidents Under NEPA (DOE, 2002)

    Broader source: Energy.gov [DOE]

    This DOE guidance clarifies and supplements "Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements." It focuses on principles of accident analyses under NEPA.

  10. Sandia Energy - Waste Isolation Pilot Plant Accident Investigation...

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

    Home Energy Nuclear Energy News News & Events Research & Capabilities Systems Analysis Materials Science Computational Modeling & Simulation Waste Isolation Pilot Plant Accident...

  11. Type B Accident Investigation of the July 12, 2007, Forklift...

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

    and maintained by the site lessee, the United States Enrichment Corporation (USEC). PDF icon Type B Accident Investigation of the July 12, 2007, Forklift and Pedestrian ...

  12. Type B Accident Investigation Board Report of the September 29...

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

    PDF icon Type B Accident Investigation Board Report of the September 29, 2010, ... Enforcement Letter, Safety and Ecology Corporation - NEL-2011-04 EA-1900: Final ...

  13. Improvement of Design Codes to Account for Accident Thermal Effects...

    Office of Environmental Management (EM)

    IMPROVEMENT OF DESIGN CODES TO ACCOUNT FOR ACCIDENT THERMAL EFFECTS ON SEISMIC PERFORMANCE Amit H. Varma, Kadir Sener, Saahas Bhardwaj Purdue University Andrew Whittaker: Univ. of...

  14. Hazard Categorization and Accident Analysis Techniques for Compliance...

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

    Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports by Diane Johnson he purpose of this DOE Standard is to...

  15. Type B Accident Investigation of the October 9, 2008 Employee...

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

    of the October 9, 2008 Employee Injured when Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office Type B Accident...

  16. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect (OSTI)

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  17. Calculation notes that support accident scenario and consequence development for the steam intrusion from interfacing systems accident

    SciTech Connect (OSTI)

    Van Vleet, R.J.; Ryan, G.W.; Crowe, R.D.; Lindberg, S.E., Fluor Daniel Hanford

    1997-03-04

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report (FSAR): Steam Intrusion From Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included in the following sections to aid in the understanding of this accident scenario. Information validation forms citing assumptions that were approved for use specifically in this analysis are included in Appendix A. Copies of these forms are also on file with TWRS Project Files. Calculations performed in this document, in general, are expressed in traditional (English) units to aid understanding of the accident scenario and related parameters.

  18. Type B Accident Investigation of the March 20, 2003, Stair Installation Accident at Building 752, Sandia National Laboratories

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Karen L. Boardman, Manager, Sandia Site Office (SSO), National Nuclear Security Administration (NNSA).

  19. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  20. Accident response group (ARG) containers for recovery of damaged warheads

    SciTech Connect (OSTI)

    York, A.R. II; Hoffman, J.P.

    1993-09-01

    This report provides an overview of the containers that are currently stored at Pantex and available for use in response to an accident or for use in any other application where a sealed containment vessel and accident resistant overpack may be needed.

  1. BWR containment failure analysis during degraded-core accidents

    SciTech Connect (OSTI)

    Yue, D.D.

    1982-06-06

    This paper presents a containment failure mode analysis during a spectrum of postulated degraded core accident sequences in a typical 1000-MW(e) boiling water reactor (BWR) with a Mark-I wetwell containment. Overtemperature failure of containment electric penetration assemblies (CEPAs) has been found to be the major failure mode during such accidents.

  2. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  3. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  4. Canister storage building design basis accident analysis documentation

    SciTech Connect (OSTI)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  5. Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video

    Broader source: Energy.gov [DOE]

    This course that provides an overview of the fundamentals of accident investigation. The course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE O 225.1B, Accident Investigations.

  6. The OSHA and EPA programs on preventing chemical accidents and potential applications in the photovoltaic industry

    SciTech Connect (OSTI)

    Fthenakis, V.M.

    1996-08-01

    OSHA issued in 1992, the Process Safety Management (PSM) of Highly Hazardous Substances. This rule requires owners/operators of facilities that handle hazardous chemicals in quantities greater than the listed thresholds to establish all the elements of a PSM. EPA has issued in June 1996, the rules for a Risk Management Program which also refers to specific substances and threshold quantities. These rules are applicable to all the facilities that use or store any of 139 regulated substances at quantities ranging from 100 lb to 10,000 lb. The RMP rule covers off-site hazards, while the OSHA Process Safety Management (PSM) rule covers worker safety issues within the plant boundary. Some of the listed substances may be found in photovoltaic manufacturing facilities. This brief report presents the basic elements of these two rules and discusses their potential applicability in the photovoltaic industry.

  7. GPHS-RTG launch accident analysis for Galileo and Ulysses

    SciTech Connect (OSTI)

    Bradshaw, C.T. )

    1991-01-01

    This paper presents the safety program conducted to determine the response of the General Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) to potential launch accidents of the Space Shuttle for the Galileo and Ulysses missions. The National Aeronautics and Space Administration (NASA) provided definition of the Shuttle potential accidents and characterized the environments. The Launch Accident Scenario Evaluation Program (LASEP) was developed by GE to analyze the RTG response to these accidents. RTG detailed response to Solid Rocket Booster (SRB) fragment impacts, as well as to other types of impact, was obtained from an extensive series of hydrocode analyses. A comprehensive test program was conducted also to determine RTG response to the accident environments. The hydrocode response analyses coupled with the test data base provided the broad range response capability which was implemented in LASEP.

  8. MELCOR accident analysis for ARIES-ACT

    SciTech Connect (OSTI)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  9. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect (OSTI)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air ?helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  10. TNRC 61 - General Provisions Use and Maintenance of Public Beaches...

    Open Energy Info (EERE)

    1 - General Provisions Use and Maintenance of Public Beaches Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: TNRC 61 - General...

  11. NMS 74-1 Environmental Improvement General Provisions | Open...

    Open Energy Info (EERE)

    1 Environmental Improvement General Provisions Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: NMS 74-1 Environmental Improvement...

  12. Title 1 General Provisions Chapter 5 Common Law; General Rights...

    Open Energy Info (EERE)

    General Provisions Chapter 5 Common Law; General Rights 3 V.S.A. Section 2809 Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute:...

  13. WPN 10-15: Final Rule on Amending Eligibility Provisions to Multifamil...

    Energy Savers [EERE]

    WPN 10-15: Final Rule on Amending Eligibility Provisions to Multifamily Buildings for the Weatherization Assistance Program WPN 10-15: Final Rule on Amending Eligibility Provisions...

  14. Intellectual Property Provisions (CDSB-115) Cooperative Agreement - Special

    Energy Savers [EERE]

    Data Statute Research, Development, or Demonstration Domestic Small Business | Department of Energy CDSB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (CDSB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Domestic Small Business PDF icon CDSB-115.pdf More Documents & Publications CDSB-1003.PDF� Intellectual Property Provisions (CDLB-115)

  15. Intellectual Property Provisions (CLB-115) Cooperative Agreement Research,

    Energy Savers [EERE]

    Development, or Demonstration Large Business and Foreign Entity | Department of Energy CLB-115) Cooperative Agreement Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (CLB-115) Cooperative Agreement Research, Development, or Demonstration Large Business and Foreign Entity PDF icon CLB-115.pdf More Documents & Publications CLB-1003.PDF� CDLB-1003.PDF� Intellectual Property Provisions (CDLB-115) Cooperative Agreement -

  16. Intellectual Property Provisions (GDLB-115) Grant - Special Data Statute

    Energy Savers [EERE]

    Research, Development, or Demonstration Large Business and Foreign Entity | Department of Energy GDLB-115) Grant - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (GDLB-115) Grant - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity PDF icon GDLB-115.pdf More Documents & Publications 10 CFR 600.325 - Intellectual Property Provisions (GDLB-1003) Grant Intellectual

  17. Intellectual Property Provisions (GDSB-115) Grant - Special Data Statute

    Energy Savers [EERE]

    Research, Development, or Demonstration Domestic Small Business | Department of Energy GDSB-115) Grant - Special Data Statute Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (GDSB-115) Grant - Special Data Statute Research, Development, or Demonstration Domestic Small Business PDF icon GDSB-115.pdf More Documents & Publications GDSB-1003.PDF� Intellectual Property Provisions (CDLB-115) Cooperative Agreement - Special Data Statute

  18. Intellectual Property Provisions (GSB-115) Grant Research, Development, or

    Energy Savers [EERE]

    Demonstration Domestic Small Business | Department of Energy GSB-115) Grant Research, Development, or Demonstration Domestic Small Business Intellectual Property Provisions (GSB-115) Grant Research, Development, or Demonstration Domestic Small Business PDF icon GSB-115.pdf More Documents & Publications GSB-1003.PDF� Intellectual Property Provisions (CLB-115) Cooperative Agreement Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property

  19. Dynamic Provisioning for Terascale Science Applications Using Hybrid

    Office of Scientific and Technical Information (OSTI)

    Circuit/Packet Technologies and 100G Transmission Systems (Technical Report) | SciTech Connect Technical Report: Dynamic Provisioning for Terascale Science Applications Using Hybrid Circuit/Packet Technologies and 100G Transmission Systems Citation Details In-Document Search Title: Dynamic Provisioning for Terascale Science Applications Using Hybrid Circuit/Packet Technologies and 100G Transmission Systems Authors: Mukherjee, Biswanath [1] + Show Author Affiliations UC, Davis Publication

  20. Final Rule on Amending Eligibility Provisions to Multifamily Buildings for

    Office of Environmental Management (EM)

    the Weatherization Assistance Program | Department of Energy Final Rule on Amending Eligibility Provisions to Multifamily Buildings for the Weatherization Assistance Program Final Rule on Amending Eligibility Provisions to Multifamily Buildings for the Weatherization Assistance Program U.S. Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy (EERE) Weatherization Assistance Program (WAP) Program Guidance 10-14 dealing with HUD multifamily buildings eligibility for

  1. Microsoft Word - Intellectual Property Provisions _NRD-1003_.doc

    Office of Environmental Management (EM)

    Intellectual Property Provisions (NRD-1003) Nonresearch and Development Nonprofit organizations are subject to the intellectual property requirements at 10 CFR 600.136(a), (c) and (d). For all other organizations, the following intellectual property provisions shall apply: (a) Recipients may copyright any work that is subject to copyright and was developed, or for which ownership was purchased, under an award. DOE reserves a royalty- free, nonexclusive and irrevocable right to reproduce, publish

  2. Calculation notes in support of TWRS FSAR spray leak accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document contains the detailed calculations that support the spray leak accident analysis in the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR). The consequence analyses in this document form the basis for the selection of controls to mitigate or prevent spray leaks throughout TWRS. Pressurized spray leaks can occur due to a breach in containment barriers along transfer routes, during waste transfers. Spray leaks are of particular safety concern because, depending on leak dimensions, and waste pressure, they can be relatively efficient generators of dispersible sized aerosols that can transport downwind to onsite and offsite receptors. Waste is transferred between storage tanks and between processing facilities and storage tanks in TWRS through a system of buried transfer lines. Pumps for transferring waste and jumpers and valves for rerouting waste are located inside below grade pits and structures that are normally covered. Pressurized spray leaks can emanate to the atmosphere due to breaches in waste transfer associated equipment inside these structures should the structures be uncovered at the time of the leak. Pressurized spray leaks can develop through holes or cracks in transfer piping, valve bodies or pump casings caused by such mechanisms as corrosion, erosion, thermal stress, or water hammer. Leaks through degraded valve packing, jumper gaskets, or pump seals can also result in pressurized spray releases. Mechanisms that can degrade seals, packing and gaskets include aging, radiation hardening, thermal stress, etc. An1782other common cause for spray leaks inside transfer enclosures are misaligned jumpers caused by human error. A spray leak inside a DST valve pit during a transfer of aging waste was selected as the bounding, representative accident for detailed analysis. Sections 2 through 5 below develop this representative accident using the DOE- STD-3009 format. Sections 2 describes the unmitigated and mitigated accident scenarios evaluated to determine the need for safety class SSCs or TSR controls. Section 3 develops the source terms associated with the unmitigated and mitigated accident scenarios. Section 4 estimates the radiological and toxicological consequences for the unmitigated and mitigated scenarios. Section 5 compares the radiological and toxicological consequences against the TWRS evaluation guidelines. Section 6 extrapolates from the representative accident case to other represented spray leak sites to assess the conservatism in using the representative case to define controls for other postulated spray leak sites throughout TWRS. Section 7 discusses the sensitivities of the consequence analyses to the key parameters and assumptions used in the analyses. Conclusions are drawn in Section 8. The analyses herein pertain to spray leaks initiated due to internal mechanisms (e.g., corrosion, erosion, thermal stress, etc). External initiators of spray leaks (e.g., excavation accidents), and natural phenomena initiators (e.g., seismic events) are to be covered in separate accident analyses.

  3. Pollution prevention efforts recognized

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

    Stories » Pollution prevention efforts recognized Pollution prevention efforts recognized Pollution prevention awards recognize individuals or teams whose efforts minimize waste, conserve resources and apply sustainable practices. April 17, 2012 George Rael presenting a bronze award for "green" purchasing to Laboratory Deputy Director Beth Sellers. George Rael, assistant manager for national security missions for the Department of Energy's Los Alamos Site Office, presents a bronze

  4. Sandia National Laboratories: Pollution Prevention

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

    Pollution Prevention Environmental Management System Pollution Prevention Sustainable Acquisition Electronics Stewardship Recycling Reuse Outreach Awards News Information...

  5. The Accident at Fukushima: What Happened?

    SciTech Connect (OSTI)

    Fujie, Takao

    2012-07-01

    At 2:46 PM, on the coast of the Pacific Ocean in eastern Japan, people were spending an ordinary afternoon. The earthquake had a magnitude of 9.0, the fourth largest ever recorded in the world. Avery large number of aftershocks were felt after the initial earthquake. More than 100 of them had a magnitude of over 6.0. There were very few injured or dead at this point. The large earthquake caused by this enormous crustal deformation spawned a rare and enormous tsunami that crashed down 30-40 minutes later. It easily cleared the high levees, washing away cars and houses and swallowing buildings of up to three stories in height. The largest tsunami reading taken from all regions was 40 meters in height. This tsunami reached the West Coast of the United States and the Pacific coast of South America, with wave heights of over two meters. It was due to this tsunami that the disaster became one of a not imaginable scale, which saw the number of dead or missing reach about 20,000 persons. The enormous tsunami headed for 15 nuclear power plants on the Pacific coast, but 11 power plants withstood the tsunami and attained cold shutdown. The flood height of the tsunami that struck each power station ranged to a maximum of 15 meters. The Fukushima Daiichi Nuclear Power Plant Units experienced the largest and the cores of three reactors suffered meltdown. As a result, more than 160,000 residents were forced to evacuate, and are still living in temporary accommodation. The main focus of this presentation is on what happened at the Fukushima Daiichi, and how station personnel responded to the accident, with considerable international support. A year after the Fukushima Daiichi accident, Japan is in the process of leveraging the lessons learned from the accident to further improve the safety of nuclear power facilities and regain the trust of society. In this connection, not only international organizations, including IAEA, and WANO, but also governmental organizations and nuclear industry representatives from various countries, have been evaluating what happened at Fukushima Daiichi. Support from many countries has contributed to successfully stabilizing the Fukushima Daiichi Nuclear Power Station. International cooperation is required as Japan started along the long road to decommissioning the reactors. Such cooperation with the international community would achieve the decommissioning of the damaged reactors. Finally, recovery plans by the Japanese government to decontaminate surrounding regions have been started in order to get residents back to their homes as early as possible. Looking at the world's nuclear power industry, there are currently approximately 440 reactors in operation and 60 under construction. Despite the dramatic consequences of the Fukushima Daiichi catastrophe it is expected that the importance of nuclear power generation will not change in the years to come. Newly accumulated knowledge and capabilities must be passed on to the next generation. This is the duty put upon us and which is one that we must embrace.

  6. Accident investigation board report on the May 14, 1997, chemical explosion at the Plutonium Reclamation Facility, Hanford Site,Richland, Washington - final report

    SciTech Connect (OSTI)

    Gerton, R.E.

    1997-07-25

    On May 14, 1997, at 7:53 p.m. (PDT), a chemical explosion occur-red in Tank A- 109 in Room 40 of the Plutonium Reclamation Facility (Facility) located in the 200 West Area of the Hanford Site, approximately 30 miles north of Richland, Washington. The inactive processing Facility is part of the Plutonium Finishing Plant (PFP). On May 16, 1997, Lloyd L. Piper, Deputy Manager, acting for John D. Wagoner, Manager, U.S. Department of Energy (DOE), Richland Operations Office (RL), formally established an Accident Investigation Board (Board) to investigate the explosion in accordance with DOE Order 225. 1, Accident Investigations. The Board commenced its investigation on May 15, 1997, completed the investigation on July 2, 1997, and submitted its findings to the RL Manager on July 26, 1997. The scope of the Board`s investigation was to review and analyze the circumstances of the events that led to the explosion; to analyze facts and to determine the causes of the accident; and to develop conclusions and judgments of need that may help prevent a recurrence of the accident. The scope also included the application of lessons learned from similar accidents within DOE. In addition to this detailed report, a companion document has also been prepared that provides a concise summary of the facts and conclusions of this report, with an emphasis on management issues (DOE/RL-97-63).

  7. Material selection for accident tolerant fuel cladding

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ≄1200°C for short (≀4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. Therefore, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steammore » and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.« less

  8. Material selection for accident tolerant fuel cladding

    SciTech Connect (OSTI)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200°C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. Therefore, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ?’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.

  9. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Chanin, D.I. ); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  10. Type B Accident Investigation At Washington Closure Hanford,...

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

    Type B Accident Investigation At Washington Closure Hanford, LLC, Employee Fall Injury on ... July 30, 2009 During D4 project demolition preparation work on the morning of July 1, ...

  11. Accidents and Intentional Destructive Acts Guidance and Requirements

    Broader source: Energy.gov [DOE]

    Accidents, as they relate to public and occupational health issues, include the determination of potential adverse effects on human health. The effects of Intentional Destructive Acts (IDAs), more...

  12. Type B Accident Investigation Of The February 25, 2009 Injury...

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

    New Mexico Type B Accident Investigation Of The February 25, 2009 Injury To A Passenger In An Electric Cart At The Waste Isolation Pilot Plant, Carlsbad, New Mexico April 1, ...

  13. Type B Accident Investigation Board Report on the Head Injury...

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

    New Mexico - August 25, 2004 Type B Accident Investigation Board Report on the Head Injury to a Miner at the Waste Isolation Pilot Plant, Carlsbad, New Mexico - August 25, ...

  14. Type B Accident Investigation Board Report of the Bechtel Jacobs...

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

    Bechtel Jacobs Company, LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, ... Type B Accident Investigation Board Report of the Bechtel Jacobs Company, LLC Employee ...

  15. Type B Accident Investigation Board Report of the Brookhaven...

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

    2009, At The 336 Building, Hanford Site, Washington Type B Accident Investigation of the Savannah River Site Arc Flash Burn Injury on September 23, 2009, in the D Area Powerhouse...

  16. Type B Accident Investigation Board Report BNFL, Inc. Employee...

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

    Park Building K-31 Type B Accident Investigation Board Report BNFL, Inc. Employee Foot Injury on December 17, 2003, at the East Tennessee Technology Park Building K-31 February 1, ...

  17. Level 1 Accident Investigation Report of August 17, 2004, Fatal...

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

    on the Grand Coulee-Bell 6 500-kV line between tower 842 and BPA's Bell Substation in Mead, Washington. (See Appendix 7, Site Map.) PDF icon Level 1 Accident Investigation Report ...

  18. Type B Accident Investigation Board Report on the October 8,...

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

    fell from a Toro Workman 3200 Utility Vehicle and fracturedhis right leg above the ankle. PDF icon Type B Accident Investigation Board Report on the October 8, 2004, Grounds...

  19. Type B Accident Investigation on the June 27, 2002, Exothermic...

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

    on the June 27, 2002, Exothermic Metal Reaction Event During Converter Disassembly in Building K-33 at the East Tennessee Technology Park Type B Accident Investigation on the June ...

  20. Type B Accident Investigation of the January 10, 2006, Flash...

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

    January 10, 2006, Flash Fire and Injury at the Savannah River National Laboratory Type B Accident Investigation of the January 10, 2006, Flash Fire and Injury at the Savannah River...

  1. Accident Investigation of the February 5, 2014, Underground Salt...

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

    5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the...

  2. Type B Accident Investigation Board Report Employee Puncture...

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

    of Energy (DOE) Savannah River Site (SRS) F-TRU Wste Facility located in the F Canyon Facility. PDF icon TYPE B ACCIDENT INVESTIGATION BOARD REPORT EMPLOYEE PUNCTURE WOUND...

  3. Core coolability following loss-of-heat sink accidents. [LMFBR

    SciTech Connect (OSTI)

    Khatib-Rahbar, M.

    1983-01-01

    Most investigations of core meltdown scenarios in liquid metal fast breeder reactors (LMFBRs) have focused on accidents resulting from unprotected transients. In comparison, protected accidents which may lead to loss of core coolability and subsequent meltdown have received considerably less attention until recently. The sequence of events leading to the protected loss-of-heat sink (LOHS) accident is among other things dependent on plant type and design. The situation is vastly different in pool-type LMFBRs as compared to the loop-type design; this is as a result of major differences in the primary system configuration, coolant inventory and the structural design. The principal aim of the present paper is to address LOHS accidents in a loop-type LMFBR in regard to physical sequences of events which could lead to loss-of-core coolability and subsequent meltdown.

  4. Accident Investigation Reports - Type B | Department of Energy

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

    independent product of the Type B Accident Investigation Board appointed by James M. Turner, Ph.D., Manager of the U.S. Department of Energy, Oakland Operations Office. July 7,...

  5. Type B Accident Investigation Board Report Grout Injection Operator...

    Energy Savers [EERE]

    and no damage to any structures inside the calvareum (i.e., no evidence of brain injury). Page 16 2.4. Investigation Readiness and Accident Scene Preservation The...

  6. Corrective Action Plan Addressing the Accident Investigation Report

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

    Corrective Action Plan Addressing the Accident Investigation Report of the February 5, 2014 Fire Event and the February 14, 2014 Radiological Release Event, Rev 1 Page 2 of 89 Table of Contents 1 Purpose ................................................................................................................................................................................................ 7 2 Summary of the

  7. Type B Accident Investigation of the Acid Vapor Inhalation on...

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

    Acid Vapor Inhalation on June 7, 2005, in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory Type B Accident Investigation of the Acid Vapor Inhalation on June 7,...

  8. Microsoft Word - ADR-Provisions-FederalAcquisitionReg.doc | Department of

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

    Energy ADR-Provisions-FederalAcquisitionReg.doc Microsoft Word - ADR-Provisions-FederalAcquisitionReg.doc PDF icon Microsoft Word - ADR-Provisions-FederalAcquisitionReg.doc More Documents & Publications ADR Provisions in Federal Acquisition Regulation (FAR) Microsoft Word - ACQUISITION LETTER.doc Microsoft Word - ADR Revised Policy82508Reformatted.doc

  9. Accident Investigation of the December 11, 2013, Integrated Device Fireset

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

    and Detonator Accidental Discharge at the Sandia National Laboratory Site 9920, Albuquerque, NM | Department of Energy December 11, 2013, Integrated Device Fireset and Detonator Accidental Discharge at the Sandia National Laboratory Site 9920, Albuquerque, NM Accident Investigation of the December 11, 2013, Integrated Device Fireset and Detonator Accidental Discharge at the Sandia National Laboratory Site 9920, Albuquerque, NM March 16, 2014 Accident Investigation of the December 11, 2013,

  10. Severe accident progression perspectives based on IPE results

    SciTech Connect (OSTI)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T.; Drouin, M.

    1996-08-01

    Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here.

  11. Volume II - Accident and Operational Safety Analysis Handbook

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

    208-2012 July 2012 DOE HANDBOOK Accident and Operational Safety Analysis Volume II: Operational Safety Analysis Techniques U.S. Department of Energy Washington, D.C. 20585 NOT MEASUREMENT SENSITIVE DOE-HDBK-1208-2012 i ACKNOWLEDGEMENTS This Department of Energy (DOE) Accident and Operational Safety Analysis Handbook was prepared under the sponsorship of the DOE Office of Health Safety and Security (HSS), Office of Corporate Safety Programs, and the Energy Facility Contractors Operating Group

  12. DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL

    Office of Environmental Management (EM)

    Video User' s Guide DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL DISCLAIMER Viewing this video and completing the enclosed printed study material do not by themselves provide sufficient skills to safely engage in or perform duties related to emergency response to a transportation accident involving radioactive material. Meeting that goal is beyond the scope of this video and requires either additional specific areas of competency or more hours of training

  13. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp

  14. Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications

    SciTech Connect (OSTI)

    VINCENT, ANDREW

    2005-04-25

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (''Quality Assurance for Safety-Related Software'') identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls to prevent or mitigate potential accidents. Over the last year, DOE has begun several processes and programs as part of the Implementation Plan commitments, and in particular, has made significant progress in addressing several sets of issues particularly important in the application of software for performing hazard and accident analysis. The work discussed here demonstrates that through these actions, Software Quality Assurance (SQA) guidance and software tools are available that can be used to improve resulting safety analysis. Specifically, five of the primary actions corresponding to the commitments made in the Implementation Plan to Recommendation 2002-1 are identified and discussed in this paper. Included are the web-based DOE SQA Knowledge Portal and the Central Registry, guidance and gap analysis reports, electronic bulletin board and discussion forum, and a DOE safety software guide. These SQA products can benefit DOE safety contractors in the development of hazard and accident analysis by precluding inappropriate software applications and utilizing best practices when incorporating software results to safety basis documentation. The improvement actions discussed here mark a beginning to establishing stronger, standard-compliant programs, practices, and processes in SQA among safety software users, managers, and reviewers throughout the DOE Complex. Additional effort is needed, however, particularly in: (1) processes to add new software applications to the DOE Safety Software Toolbox; (2) improving the effectiveness of software issue communication; and (3) promoting a safety software quality assurance culture.

  15. Type B Accident Investigation Board Report on the March 27, 1998, Rotating Shaft Accident at the Ames Laboratory, Ames, Iowa

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by John Kennedy, Acting Manager, Chicago Operations Office, U.S. Department of Energy (DOE).

  16. Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    Safety after a Traffic Accident to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Twitter Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Google Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Delicious Rank Alternative Fuels Data Center: Natural Gas Safety after a

  17. Energy Department Seeks Public Comment on Standby Support Provisions of

    Energy Savers [EERE]

    Energy Policy Act of 2005 | Department of Energy Public Comment on Standby Support Provisions of Energy Policy Act of 2005 Energy Department Seeks Public Comment on Standby Support Provisions of Energy Policy Act of 2005 November 17, 2005 - 2:49pm Addthis Workshop is first step to develop regulation for risk insurance for nuclear plants WASHINGTON, DC -- The U.S. Department of Energy (DOE) announced today that a workshop to receive public comments on the "standby support"

  18. SILENE Benchmark Critical Experiments for Criticality Accident Alarm Systems

    SciTech Connect (OSTI)

    Miller, Thomas Martin; Reynolds, Kevin H.

    2011-01-01

    In October 2010 a series of benchmark experiments was conducted at the Commissariat a Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE [1] facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems (CAASs). This presentation will discuss the geometric configuration of these experiments and the quantities that were measured and will present some preliminary comparisons between the measured data and calculations. This series consisted of three single-pulsed experiments with the SILENE reactor. During the first experiment the reactor was bare (unshielded), but during the second and third experiments it was shielded by lead and polyethylene, respectively. During each experiment several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor, and some of these detectors were themselves shielded from the reactor by high-density magnetite and barite concrete, standard concrete, and/or BoroBond. All the concrete was provided by CEA Saclay, and the BoroBond was provided by Y-12 National Security Complex. Figure 1 is a picture of the SILENE reactor cell configured for pulse 1. Also included in these experiments were measurements of the neutron and photon spectra with two BICRON BC-501A liquid scintillators. These two detectors were provided and operated by CEA Valduc. They were set up just outside the SILENE reactor cell with additional lead shielding to prevent the detectors from being saturated. The final detectors involved in the experiments were two different types of CAAS detectors. The Babcock International Group provided three CIDAS CAAS detectors, which measured photon dose and dose rate with a Geiger-Mueller tube. CIDAS detectors are currently in use at Y-12 in the newly constructed Highly Enriched Uranium Materials Facility. The second CAAS detector used a {sup 6}LiF TLD to absorb neutrons and a silicon detector to count the charge particles released by these absorption events. Lawrence Livermore National Laboratory provided four of these detectors, which had formerly been used at the Rocky Flats facility in the United States.

  19. Intellectual Property Provisions (GLB-115) Grant Research, Development, or

    Energy Savers [EERE]

    Demonstration Large Business and Foreign Entity | Department of Energy GLB-115) Grant Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (GLB-115) Grant Research, Development, or Demonstration Large Business and Foreign Entity PDF icon GLB-115.pdf More Documents & Publications GLB-1003.PDF� CLB-1003.PDF� CDLB

  20. Intellectual Property Provisions (GNP-115) Grant and Cooperative Agreement

    Energy Savers [EERE]

    Research, Development, or Demonstration Non-Federal Entity (State, Local government, Indian tribe, Institution of higher education, or Nonprofit organization) | Department of Energy GNP-115) Grant and Cooperative Agreement Research, Development, or Demonstration Non-Federal Entity (State, Local government, Indian tribe, Institution of higher education, or Nonprofit organization) Intellectual Property Provisions (GNP-115) Grant and Cooperative Agreement Research, Development, or Demonstration

  1. Analyses of Selected Provisions of Proposed Energy Legislation: 2003

    Reports and Publications (EIA)

    2003-01-01

    This study responds to a July 31, 2003 request from Senator Byron L. Dorgan. The study is based primarily on analyses the Energy Information Administration has previously done for studies requested by Congress. It includes analysis of the Renewable Portfolio Standard, Renewable Fuels Standard, production in the Alaskan National Wildlife Refuge, the construction of an Alaskan Natural Gas pipeline, and various tax provisions.

  2. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect (OSTI)

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K.

    2012-07-01

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  3. ATWS at Browns Ferry Unit One - accident sequence analysis

    SciTech Connect (OSTI)

    Harrington, R.M.; Hodge, S.A.

    1984-07-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence the quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.

  4. DOE - NNSA/NFO -- News & Views Accident Trap

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

    Accident Traps Four Workers 1,800 Feet below Ground Photo - rescue from 1,800 feet below Over the past 42 years the Nevada Test Site has earned an excellent safety record. Thousands of workers have completed millions of accident-free hours at this heavy industry site. Since 1957 (no accurate records exist for 1951-56) there have been 46 fatalities on the Test Site, and six fatalities at other Test-Site-related locations. Being safety conscious and prepared minimizes the impact of the occasional

  5. Intellectual Property Provisions (CDLB-115) Cooperative Agreement - Special

    Energy Savers [EERE]

    Data Statute Research, Development, or Demonstration Large Business and Foreign Entity | Department of Energy CDLB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity Intellectual Property Provisions (CDLB-115) Cooperative Agreement - Special Data Statute Research, Development, or Demonstration Large Business and Foreign Entity PDF icon CDLB-115.pdf More Documents & Publications CDLB-1003.PDF� Intellectual

  6. Intellectual Property Provisions (CSB-1003) Cooperative Agreement Research,

    Energy Savers [EERE]

    Development, or Demonstration Domestic Small Businesses | Department of Energy 003) Cooperative Agreement Research, Development, or Demonstration Domestic Small Businesses Intellectual Property Provisions (CSB-1003) Cooperative Agreement Research, Development, or Demonstration Domestic Small Businesses 01. FAR 52.227-1 Authorization and Consent (JUL 1995)-Alternate I (APR 1984) 02. FAR 52.227-2 Notice and Assistance Regarding Patent and Copyright Infringement (AUG 1996) This clause is not

  7. EPACT2005: Status of Provisions (Update) (released in AEO2007)

    Reports and Publications (EIA)

    2007-01-01

    The Energy Policy Act 2005 (EPACT) was signed into law by President Bush on August 8, 2005, and became Public Law 109-058. A number of provisions from EPACT2005 were included in the Annual Energy Outlook 2006 (AEO) projections. Many others were not considered in AEO2006particularly, those that require funding appropriations or further specification by federal agencies or Congress before implementation.

  8. Intellectual Property Provisions - Cooperative Agreement - Special Data Statute

    Office of Environmental Management (EM)

    15 Intellectual Property Provisions (CSB-115) Cooperative Agreement Research, Development, or Demonstration Domestic Small Business 01. FAR 52.227-1 Authorization and Consent (DEC 2007) Alternate I (APR 1984) 02. FAR 52.227-2 Notice and Assistance Regarding Patent and Copyright Infringement (DEC 2007) 03. 2 CFR 910 Appendix A of Subpart D Rights in Data - General 04. 2 CFR 910 Appendix A of Subpart D Patent Rights (Small Business Firms and Nonprofit Organizations) NOTE: In reading these

  9. U-161: Citrix Provisioning Services Unspecified Flaw Lets Remote Users Execute Arbitrary Code

    Broader source: Energy.gov [DOE]

    A vulnerability was reported in Citrix Provisioning Services. A remote user can execute arbitrary code on the target system.

  10. 10 CFR 600.325 - Intellectual Property Provisions (GDLB-1003) Grant |

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

    Department of Energy 10 CFR 600.325 - Intellectual Property Provisions (GDLB-1003) Grant 10 CFR 600.325 - Intellectual Property Provisions (GDLB-1003) Grant Special Data Statute Research, Development, or Demonstration Large Businesses, State and Local Governments, and Foreign Entities PDF icon 10 CFR 600.325 - Intellectual Property Provisions (GDLB-1003) Grant More Documents & Publications Intellectual Property Provisions (GDLB-115) Grant - Special Data Statute Research, Development, or

  11. Type A Accident Investigation Board Report on the July 1, 2008...

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

    July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration Type A Accident Investigation Board Report on the July 1, 2008, of the Vehicle Fatality ...

  12. Calculation Notes for Subsurface Leak Resulting in Pool, TWRS FSAR Accident Analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Subsurface Leaks Resulting in Pool.

  13. Evaluation of severe accident risks: Surry Unit 1

    SciTech Connect (OSTI)

    Breeding, R.J. ); Helton, J.C. ); Murfin, W.B. ); Smith, L.N. )

    1990-10-01

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Surry Power Station, Unit 1. This power plant, located in southeastern Virginia, is operated by the Virginia Electric Power Corp. The emphasis in this risk analysis was not on determining a so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiation by events, both internal to the power station and external to the power station were assessed. This document, Volume 3, Revision 1, Part 2, provides Appendices A through E to this report. These appendices contain: supporting information for the accident progression analysis; the source term analysis; the consequence analysis; risk results; and sampling information.

  14. Source terms for plutonium aerosolization from nuclear weapon accidents

    SciTech Connect (OSTI)

    Stephens, D.R.

    1995-07-01

    The source term literature was reviewed to estimate aerosolized and respirable release fractions for accidents involving plutonium in high-explosive (HE) detonation and in fuel fires. For HE detonation, all estimates are based on the total amount of Pu. For fuel fires, all estimates are based on the amount of Pu oxidized. I based my estimates for HE detonation primarily upon the results from the Roller Coaster experiment. For hydrocarbon fuel fire oxidation of plutonium, I based lower bound values on laboratory experiments which represent accident scenarios with very little turbulence and updraft of a fire. Expected values for aerosolization were obtained from the Vixen A field tests, which represent a realistic case for modest turbulence and updraft, and for respirable fractions from some laboratory experiments involving large samples of Pu. Upper bound estimates for credible accidents are based on experiments involving combustion of molten plutonium droplets. In May of 1991 the DOE Pilot Safety Study Program established a group of experts to estimate the fractions of plutonium which would be aerosolized and respirable for certain nuclear weapon accident scenarios.

  15. T-541: Citrix Provisioning Services Unspecified Flaw Let's Remote Users Execute Arbitrary Code

    Broader source: Energy.gov [DOE]

    A vulnerability has been identified in Citrix Provisioning Services that could result in arbitrary code execution. This vulnerability can be triggered by an attacker sending a specially crafted packet to the Provisioning Services server. This vulnerability is present in all supported versions of Citrix Provisioning Services up to and including version 5.6.

  16. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect (OSTI)

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

  17. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  18. Microsoft Word - 2015.06.22 - Report to Congress - Accident Tolerant...

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

    ROADMAP: DEVELOPMENT OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Development of ...... ROADMAP: DEVELOPMENT OF LWR FUELS WITH ENHANCED ...

  19. Type A Accident Investigation Board Report on the July 1, 2008, of the

    Energy Savers [EERE]

    Vehicle Fatality Accident-Western Area Power Marketing Administration | Department of Energy July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration Type A Accident Investigation Board Report on the July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration August 29, 2008 At approximately 1210 CDT, July 1, 2008, three Western Area Power Administration (Western) employees were traveling south on North Dakota gravel road 59th

  20. Type B Accident Investigation Board Report of the April 23, 1997,

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

    Helicopter Accident at Raton Pass, Raton Pass, Colorado | Department of Energy April 23, 1997, Helicopter Accident at Raton Pass, Raton Pass, Colorado Type B Accident Investigation Board Report of the April 23, 1997, Helicopter Accident at Raton Pass, Raton Pass, Colorado May 1997 On April 23, 1997, a helicopter belonging to the Western Area Power Administration (Western) crashed near the summit of Raton Pass in southern Colorado. On April 24, 1997, Michael S. Cowan, Western's Chief Program

  1. Type B Accident Investigation of the July 12, 2007, Forklift and Pedestrian

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

    Accident at the Paducah Gaseous Diffusion Plant, Portsmouth/Paducah Project Office | Department of Energy 2, 2007, Forklift and Pedestrian Accident at the Paducah Gaseous Diffusion Plant, Portsmouth/Paducah Project Office Type B Accident Investigation of the July 12, 2007, Forklift and Pedestrian Accident at the Paducah Gaseous Diffusion Plant, Portsmouth/Paducah Project Office April 14, 2008 On July 12, 2007, an employee at the Paducah Gaseous Diffusion Plant (PGDP) was walking alone during

  2. Type B Accident Investigation of the July 14, 2005, Americium Contamination

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

    Accident at the Sigma Facility, Los Alamos National Laboratory | Department of Energy 14, 2005, Americium Contamination Accident at the Sigma Facility, Los Alamos National Laboratory Type B Accident Investigation of the July 14, 2005, Americium Contamination Accident at the Sigma Facility, Los Alamos National Laboratory January 1, 2006 On July 14, 2005, a worker at the Los Alamos National Laboratory received and opened a shipment of radioactive material from another facility in the

  3. Type B Accident Investigation of the July 31, 2006, Fall from Ladder

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

    Accident at the Lawrence Livermore National Laboratory, Livermore, California | Department of Energy 31, 2006, Fall from Ladder Accident at the Lawrence Livermore National Laboratory, Livermore, California Type B Accident Investigation of the July 31, 2006, Fall from Ladder Accident at the Lawrence Livermore National Laboratory, Livermore, California October 25, 2006 Early on the morning of July 31, 2006, an electrician in the Plant Engineering (PE) Department of the Lawrence Livermore

  4. Type B Accident Investigation on the February 17, 2004, Personal Injury

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

    Accident, Bettis Atomic Power Laboratory | Department of Energy on the February 17, 2004, Personal Injury Accident, Bettis Atomic Power Laboratory Type B Accident Investigation on the February 17, 2004, Personal Injury Accident, Bettis Atomic Power Laboratory August 16, 2004 Prime contractors need to provide a safe work environment for the entire facility site, including parking lots and outdoor pedestrian walkways. Particular attention needs to be given to areas that must be traversed by

  5. Level 1 Accident Report of the March 1, 2010 Bobcat Fatality at BPA's White

    Office of Environmental Management (EM)

    Bluffs Substation | Department of Energy Report of the March 1, 2010 Bobcat Fatality at BPA's White Bluffs Substation Level 1 Accident Report of the March 1, 2010 Bobcat Fatality at BPA's White Bluffs Substation March 31, 2010 On March 2, 2010 at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was convened to investigate an accident in which a supplemental labor contractor was fatally injured in a Bobcat/backhoe accident at the

  6. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    RITTMANN, P.D.

    1999-10-07

    Three bounding accidents postdated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing, and a hydrogen explosion. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  7. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    2000-01-10

    Four bounding accidents postulated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing a hydrogen explosion, and a fire breaching filter vessel and enclosure. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  8. Microsoft Word - 2015.06.22 - Report to Congress - Accident Tolerant Fuels

    Office of Environmental Management (EM)

    ROADMAP: DEVELOPMENT OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Development of Light Water Reactor Fuels with Enhanced Accident Tolerance Report to Congress April 2015 United States Department of Energy Washington, DC 20585 _____________________________________________________________________________ ROADMAP: DEVELOPMENT OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Message from the Assistant Secretary for Nuclear Energy In the Senate Appropriations Committee Report (Senate

  9. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    SciTech Connect (OSTI)

    Robb, Kevin R

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  10. Type B Accident Investigation Board Report of the January 20, 1998, Electrical Accident at the Casa Grande Substation,South of Phoenix, Arizona

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type-B Accident Investigation Board appointed by Michael S.Cowan, Chief Program Officer, Western Area Power Administration.

  11. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect (OSTI)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to automatically respond to the change in reactor conditions and to result in a benign response to these events. This approach has the advantage of being relatively simple to implement, and does not face the issue of reliability since only fundamental physical phenomena are used in a passive manner, not active engineered systems. However, the challenge is to present a convincing case that such passive means can be implemented and used. The purpose of this paper is to describe this third approach in detail, the technical basis and experimental validation for the approach, and the resulting reactor performance that can be achieved for ATWS events.

  12. CEQ Regulations for Implementing the Procedural Provisions of NEPA

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

    Council on Environmental Quality Executive Office of the President REGULATIONS For Implementing The Procedural Provisions Of The NATIONAL ENVIRONMENTAL POLICY ACT Reprint 40 CFR Parts 1500-1508 (2005) This page is blank (inside front cover) TABLE OF CONTENTS PART 1500-PURPOSE, POLICY AND MANDATE Sec. 1500.1 Purpose. 1500.2 Policy. 1500.3 Mandate. 1500.4 Reducing paperwork. 1500.5 Reducing delay. 1500.6 Agency authority. PART 1501-NEPA AND AGENCY PLANNING Sec. 1501.1 Purpose. 1501.2 Apply NEPA

  13. Initiatives for proliferation prevention

    SciTech Connect (OSTI)

    1997-04-01

    Preventing the proliferation of weapons of mass destruction is a central part of US national security policy. A principal instrument of the Department of Energy`s (DOE`s) program for securing weapons of mass destruction technology and expertise and removing incentives for scientists, engineers and technicians in the newly independent states (NIS) of the former Soviet Union to go to rogue countries or assist terrorist groups is the Initiatives for Proliferation Prevention (IPP). IPP was initiated pursuant to the 1994 Foreign Operations Appropriations Act. IPP is a nonproliferation program with a commercialization strategy. IPP seeks to enhance US national security and to achieve nonproliferation objectives by engaging scientists, engineers and technicians from former NIS weapons institutes; redirecting their activities in cooperatively-developed, commercially viable non-weapons related projects. These projects lead to commercial and economic benefits for both the NIS and the US IPP projects are funded in Russian, Ukraine, Kazakhstan and Belarus. This booklet offers an overview of the IPP program as well as a sampling of some of the projects which are currently underway.

  14. The Nevada railroad system: Physical, operational, and accident characteristics

    SciTech Connect (OSTI)

    1991-09-01

    This report provides a description of the operational and physical characteristics of the Nevada railroad system. To understand the dynamics of the rail system, one must consider the system`s physical characteristics, routing, uses, interactions with other systems, and unique operational characteristics, if any. This report is presented in two parts. The first part is a narrative description of all mainlines and major branchlines of the Nevada railroad system. Each Nevada rail route is described, including the route`s physical characteristics, traffic type and volume, track conditions, and history. The second part of this study provides a more detailed analysis of Nevada railroad accident characteristics than was presented in the Preliminary Nevada Transportation Accident Characterization Study (DOE, 1990).

  15. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect (OSTI)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

  16. Input-output model for MACCS nuclear accident impacts estimationč

    SciTech Connect (OSTI)

    Outkin, Alexander V.; Bixler, Nathan E.; Vargas, Vanessa N

    2015-01-27

    Since the original economic model for MACCS was developed, better quality economic data (as well as the tools to gather and process it) and better computational capabilities have become available. The update of the economic impacts component of the MACCS legacy model will provide improved estimates of business disruptions through the use of Input-Output based economic impact estimation. This paper presents an updated MACCS model, bases on Input-Output methodology, in which economic impacts are calculated using the Regional Economic Accounting analysis tool (REAcct) created at Sandia National Laboratories. This new GDP-based model allows quick and consistent estimation of gross domestic product (GDP) losses due to nuclear power plant accidents. This paper outlines the steps taken to combine the REAcct Input-Output-based model with the MACCS code, describes the GDP loss calculation, and discusses the parameters and modeling assumptions necessary for the estimation of long-term effects of nuclear power plant accidents.

  17. Accident Investigation of the October 1, 2013, Tice Electric Company

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

    Employee Fatality near Patrick's Knob Radio Station, Bonneville Power Administration | Department of Energy October 1, 2013, Tice Electric Company Employee Fatality near Patrick's Knob Radio Station, Bonneville Power Administration Accident Investigation of the October 1, 2013, Tice Electric Company Employee Fatality near Patrick's Knob Radio Station, Bonneville Power Administration November 22, 2013 On October 2, 2013, at the request of the Bonneville Power Administration (BPA) Chief Safety

  18. In a mining accident, first responders are working against

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

    a mining accident, first responders are working against the clock and against a myriad of dangers such as debris, poisonous gases, flooding, explosive vapors, and unstable structures to assess the situation and rescue trapped miners. These unknown and potentially deadly conditions create a challenge for first responders and often limit their ability to assess the situation and respond in a timely matter. There is a need for a robotic system that could be used to support a mine rescue team,

  19. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect (OSTI)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tesmer, Joseph R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  20. Type B Accident Investigation At Washington Closure Hanford, LLC, Employee

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

    Fall Injury on July 1, 2009, At The 336 Building, Hanford Site, Washington | Department of Energy At Washington Closure Hanford, LLC, Employee Fall Injury on July 1, 2009, At The 336 Building, Hanford Site, Washington Type B Accident Investigation At Washington Closure Hanford, LLC, Employee Fall Injury on July 1, 2009, At The 336 Building, Hanford Site, Washington July 30, 2009 During D4 project demolition preparation work on the morning of July 1, 2009, in Hanford's 300 Area, a millwright

  1. Type B Accident Investigation Board Report Subcontractor Radioactive

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

    Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) | Department of Energy Subcontractor Radioactive Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) Type B Accident Investigation Board Report Subcontractor Radioactive Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) August 17, 2004 On Friday,

  2. Type B Accident Investigation Board Report of the Brookhaven National

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

    Laboratory Employee Injury at Building 1005H on October 9, 2009 | Department of Energy of the Brookhaven National Laboratory Employee Injury at Building 1005H on October 9, 2009 Type B Accident Investigation Board Report of the Brookhaven National Laboratory Employee Injury at Building 1005H on October 9, 2009 December 11, 2009 On the afternoon of October 9, 2009, a Lead Rigger for Brookhaven Science Associates (BSA), LLC at the Brookhaven National laboratory (BNL) wasinjured while at the

  3. Type B Accident Investigation Report on the Exertional Heat Illnesses

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

    during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July 13, 2006 | Department of Energy on the Exertional Heat Illnesses during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July 13, 2006 Type B Accident Investigation Report on the Exertional Heat Illnesses during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July 13, 2006 July 13, 2006 This Report addresses three injuries that occurred on June 15, 2006 during the

  4. Accident Investigation Report - Radiological Release | Department of Energy

    Energy Savers [EERE]

    Radiological Release Accident Investigation Report - Radiological Release On February 14, 2014, an airborne radiological release occurred at the Department of Energy Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. Because access to the underground was restricted following the event, the investigation was broken into two phases. The Phase 1 report focused on how the radiological material was released into the atmosphere and Phase 2, performed once limited access to the underground

  5. KERENA safety concept in the context of the Fukushima accident

    SciTech Connect (OSTI)

    Zacharias, T.; Novotny, C.; Bielor, E.

    2012-07-01

    Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basic physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)

  6. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    SciTech Connect (OSTI)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  7. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

    Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  8. Waste Minimization and Pollution Prevention

    Broader source: Energy.gov [DOE]

    The team supports efforts that promote a more sustainable environment and implements pollution prevention activities, as deemed appropriate for LM operations and approved by LM, as defined in:

  9. Guidelines for Provision and Interchange of Geothermal Data Assets

    SciTech Connect (OSTI)

    none,

    2014-07-03

    The US Department of Energy Office of Geothermal Technologies (OGT) is funding and overseeing the development of the National Geothermal Data System (NGDS), a distributed information system providing access to integrated data in support of, and generated in, all phases of geothermal development. NGDS is being built in an open paradigm and will employ state-of-the-art informatics approaches and capabilities to advance the state of geothermal knowledge in the US. This document presents guidelines related to provision and interchange of data assets in the context of the National Geothermal Data System. It identifies general specifications for NGDS catalog metadata and data content, and provides specific instructions for preparation and submission of data assets by OGT-funded projects.

  10. Impact of ASCE-95 wind provisions on reinforced concrete chimneys

    SciTech Connect (OSTI)

    Joshi, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Porthouse, R.A. [Chimney Consultants, Inc., West Lebanon, NH (United States)

    1999-11-01

    This paper presents a summary of the study on the impact of the ASCE 7-95 provisions primarily on the along wind load calculations for reinforced concrete chimneys. The results of this study provided, in part, a basis for the adoption of the revised load and strength factors for wind load design in ACI 307-98. The study was undertaken to: compare unfactored along wind chimney loads resulting from fastest miles per hour (FMPH) and three second gust (TSG) procedures; quantify design changes resulting from FMPH to TSG in terms of the reinforcing steel; review the historical basis for the current load factors (LF) and strength factors (SF); and to recommend a set of LF and SF for the design of chimneys without significantly changing the present design confidence level.

  11. Impact of Renewable Fuels Standard/MTBE Provisions of S. 517 Requested by Sens. Daschle & Murkowski

    Reports and Publications (EIA)

    2002-01-01

    Additional analysis of the impact of the Renewable Fuels Standard (RFS) and methyl tertiary butyl ether (MTBE) ban provisions of S. 517.

  12. NMAC 19.14.1 Geothermal Power General Provisions | Open Energy...

    Open Energy Info (EERE)

    Geothermal Power General ProvisionsLegal Abstract These regulations outline the procedures for dealing with geothermal power issues in New Mexico. These rules are designed to...

  13. Montana MCA 75-20-104, Policy and General Provisions Definitions...

    Open Energy Info (EERE)

    75-20-104, Policy and General Provisions Definitions for Major Facility Siting Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document-...

  14. Implantation, Activation, Characterization and Prevention/Mitigation...

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

    Activation, Characterization and PreventionMitigation of Internal Short Circuits in Lithium-Ion Cells Implantation, Activation, Characterization and Prevention...

  15. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect (OSTI)

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  16. Markov Model of Accident Progression at Fukushima Daiichi

    SciTech Connect (OSTI)

    Cuadra A.; Bari R.; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G.; Mubayi, V.; Pratt, T.; Yue, M.

    2012-11-11

    On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

  17. Safety evaluation of MHTGR licensing basis accident scenarios

    SciTech Connect (OSTI)

    Kroeger, P.G.

    1989-04-01

    The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential accident scenarios that might lead to fuel failures due to excessive core temperatures and/or to vessel damage, due to excessive vessel temperatures. The design basis accident scenario leading to the highest vessel temperatures is the depressurized core heatup scenario without any forced cooling and with decay heat rejection to the passive Reactor Cavity Cooling System (RCCS). This scenario was evaluated, including numerous parametric variations of input parameters, like material properties and decay heat. It was found that significant safety margins exist, but that high confidence levels in the core effective thermal conductivity, the reactor vessel and RCCS thermal emissivities and the decay heat function are required to maintain this safety margin. Severe accident extensions of this depressurized core heatup scenario included the cases of complete RCCS failure, cases of massive air ingress, core heatup without scram and cases of degraded RCCS performance due to absorbing gases in the reactor cavity. Except for no-scram scenarios extending beyond 100 hr, the fuel never reached the limiting temperature of 1600/degree/C, below which measurable fuel failures are not expected. In some of the scenarios, excessive vessel and concrete temperatures could lead to investment losses but are not expected to lead to any source term beyond that from the circulating inventory. 19 refs., 56 figs., 11 tabs.

  18. Temperature of aircraft cargo flame exposure during accidents involving fuel spills

    SciTech Connect (OSTI)

    Mansfield, J.A.

    1993-01-01

    This report describes an evaluation of flame exposure temperatures of weapons contained in alert (parked) bombers due to accidents that involve aircraft fuel fires. The evaluation includes two types of accident, collisions into an alert aircraft by an aircraft that is on landing or take-off, and engine start accidents. Both the B-1B and B-52 alert aircraft are included in the evaluation.

  19. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect (OSTI)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  20. Accident Investigation of the June 1, 2013, Stairway Fall Resulting in a

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

    Federal Employee Fatality at DOE Headquarters Germantown, Maryland | Department of Energy , 2013, Stairway Fall Resulting in a Federal Employee Fatality at DOE Headquarters Germantown, Maryland Accident Investigation of the June 1, 2013, Stairway Fall Resulting in a Federal Employee Fatality at DOE Headquarters Germantown, Maryland On June 28, 2013, an Accident Investigation Board was appointed to investigate an accident at the Department of Energy Germantown Headquarters facility, on June

  1. Type B Accident Investigation of the Arc Flash at Brookhaven National

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

    Laboratory, April 14, 2006 | Department of Energy Arc Flash at Brookhaven National Laboratory, April 14, 2006 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 February 10, 2006 An accident at the Idaho National Laboratory (INL) was investigated in which a technician sustained a serious injury to his right hand while operating a table saw. In conducting its investigation, the Accident Investigation Board (the Board) used various analytical

  2. Analysis of Three Mile Island-Unit 2 accident

    SciTech Connect (OSTI)

    Not Available

    1980-03-01

    The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert.

  3. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that significantly exceeded QE limits for extended time periods for the low frequency STSBO sequence evaluated in this study. It is recognized that the core damage frequency (CDF) of the sequence evaluated in this scoping effort would be considerably lower if evaluations considered new FLEX equipment being installed by industry. Nevertheless, because of uncertainties in instrumentation response when exposed to conditions beyond QE limits and alternate challenges associated with different sequences that may impact sensor performance, it is recommended that additional evaluations of instrumentation performance be completed to provide confidence that operators have access to accurate, relevant, and timely information on the status of reactor systems for a broad range of challenges associated with risk important severe accident sequences.

  4. Next-generation nuclear fuel withstands high-temperature accident

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

    conditions U.S. DEPARTMENT OF ENERGY IDAHO FALLS, IDAHO, 83403 For Immediate Release: Sept. 25, 2013 Media Contacts: Teri Ehresman, 208-526-7785 teri.ehresman@inl.gov Bill Cabage (ORNL), 865-574-4399, cabagewh@ornl.gov Next-generation nuclear fuel withstands high-temperature accident conditions IDAHO FALLS - A safer and more efficient nuclear fuel is on the horizon. A team of researchers at the U.S. Department of Energy's Idaho National Laboratory (INL) and Oak Ridge National Laboratory

  5. Advance plant severe accident/thermal hydraulic issues for ACRS

    SciTech Connect (OSTI)

    Kress, T.S.

    1994-09-01

    The ACRS has been reviewing various advance plant designs for certification. The most active reviews have been for the ABWR, AP600, and System 80+. We have completed the reviews for ABWR and System 80+ and are presently concentrating on AP600. The ACRS gave essentially unqualified certification approval for the two completed reviews, yet,,during the process of review a number of issues arose and the plant designs changed somewhat to accommodate some of the ACRS concerns. In this talk, I will describe some of the severe accident and thermal hydraulic related issues we discussed in our reviews.

  6. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect (OSTI)

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  7. A comparative analysis of accident risks in fossil, hydro, and nuclear energy chains

    SciTech Connect (OSTI)

    Burgherr, P.; Hirschberg, S.

    2008-07-01

    This study presents a comparative assessment of severe accident risks in the energy sector, based on the historical experience of fossil (coal, oil, natural gas, and LPG (Liquefied Petroleum Gas)) and hydro chains contained in the comprehensive Energy-related Severe Accident Database (ENSAD), as well as Probabilistic Safety Assessment (PSA) for the nuclear chain. Full energy chains were considered because accidents can take place at every stage of the chain. Comparative analyses for the years 1969-2000 included a total of 1870 severe ({>=} 5 fatalities) accidents, amounting to 81,258 fatalities. Although 79.1% of all accidents and 88.9% of associated fatalities occurred in less developed, non-OECD countries, industrialized OECD countries dominated insured losses (78.0%), reflecting their substantially higher insurance density and stricter safety regulations. Aggregated indicators and frequency-consequence (F-N) curves showed that energy-related accident risks in non-OECD countries are distinctly higher than in OECD countries. Hydropower in non-OECD countries and upstream stages within fossil energy chains are most accident-prone. Expected fatality rates are lowest for Western hydropower and nuclear power plants; however, the maximum credible consequences can be very large. Total economic damages due to severe accidents are substantial, but small when compared with natural disasters. Similarly, external costs associated with severe accidents are generally much smaller than monetized damages caused by air pollution.

  8. Type A Accident Investigation Board Report on the April 19, 1999...

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

    resulting in death. The Assistant Secretary for the Office of Environment, Safety and Health (EH) delegated the responsibility for conducting a Type A accident investigation to...

  9. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple...

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

    direct cause of the accident was the release of airborne contamination from a degraded package that contained cellulose material and plutonium-238 residues. PDF icon Type B...

  10. Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS

    Broader source: Energy.gov [DOE]

    DOE O 225.1B prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and...

  11. Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident |

    Energy Savers [EERE]

    Department of Energy Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident November 26, 2013 - 12:00pm Addthis BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. Russ Hall,

  12. Level 1 Accident Investigation Report of August 17, 2004, Fatal Aircraft

    Office of Environmental Management (EM)

    Accident on the Grand Coulee-Bell No.6, 500 kV Line | Department of Energy Investigation Report of August 17, 2004, Fatal Aircraft Accident on the Grand Coulee-Bell No.6, 500 kV Line Level 1 Accident Investigation Report of August 17, 2004, Fatal Aircraft Accident on the Grand Coulee-Bell No.6, 500 kV Line OCTOBER 1, 2004 On August 17, 2004, at approximately 0940, a Bonneville Power Administration (BPA) pilot was killed in the crash of a Bell 206BIII helicopter while stringing "sock

  13. Accident Investigation of the February 7, 2013, Scissor Lift Accident in the West Hackberry Brine Tank-14 Resulting in Injury, Strategic Petroleum Reserve West Hackberry, LA

    Broader source: Energy.gov [DOE]

    On February 15, 2013, an Accident Investigation Board (the Board) was appointed to investigate an accident that resulted in serious injuries caused when a scissor lift tipped over in Brine Tank-14 (WHT-14) at the Strategic Petroleum Reserve, West Hackberry, Louisiana, site on February 7, 2013. The Board’s responsibilities have been completed with respect to this investigation. The analysis and the identification of the direct cause, root causes, contributing causes, and judgments of need resulting from this investigation were performed in accordance with the Department of Energy (DOE) Order 225.1B, Accident Investigations.

  14. Impact of Renewable Fuels Standard/MTBE Provisions of S. 1766

    Reports and Publications (EIA)

    2002-01-01

    This service report addresses the Renewable Fuels Standard (RFS)/methyl tertiary butyl ether (MTBE) provisions of S. 1766. The 'S. 1766' Case reflects provisions of S. 1766 including a renewable fuels standard (RFS) reaching five billion gallons by 2012, a complete phase-out of MTBE within four years, and the option for states to waive the oxygen requirement for reformulated gasoline (RFG).

  15. PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

    SciTech Connect (OSTI)

    , K; Jonathan Lowrie, J; David Thoman , D; Austin Keller , A

    2008-07-30

    Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases.

  16. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  17. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    SciTech Connect (OSTI)

    Wenxi Tian; Suizheng Qiu; Guanghui Su; Dounan Jia [Xi'an Jiaotong University, 28 Xianning Road, Xi'an 710049 (China); Xingmin Liu - China Institute of Atomic Energy

    2006-07-01

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  18. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect (OSTI)

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  19. Type B Accident Investigation Board Report on the August 5, 1998, Load Haul Dump Accident at U16b Tunnel, Nevada Test Site

    Broader source: Energy.gov [DOE]

    Thisis theType B Accident Investigation Board report of an industrial accident at the Nevada Test site (NTS), U16b tunnel in which a Bechtel Nevada (BN) employee suffered a compressed skull fracture as a result of being struck onthe head by a valve and fitting assembly on the end of a hose whichhad been broken from a water pipe by a moving piece of construction equipment.

  20. Recovery sequences for a station blackout accident at the Grand Gulf Nuclear Station

    SciTech Connect (OSTI)

    Carbajo, J.J. [Martin Marietta Energy Systems, Oak Ridge, TN (United States)

    1995-12-31

    Recovery sequences for a low-pressure, short term, station blackout severe accident at the Grand Gulf power plant have been investigated using the computer code MELCOR, version 1.8.3 PN. This paper investigates the effect of reflood timing and mass flow rate on accident recovery.

  1. Shipping container response to three severe railway accident scenarios

    SciTech Connect (OSTI)

    Mok, G.C.; Fischer, L.E.; Murty, S.S.; Witte, M.C.

    1998-04-01

    The probability of damage and the potential resulting hazards are analyzed for a representative rail shipping container for three severe rail accident scenarios. The scenarios are: (1) the rupture of closure bolts and resulting opening of closure lid due to a severe impact, (2) the puncture of container by an impacting rail-car coupler, and (3) the yielding of container due to side impact on a rigid uneven surface. The analysis results indicate that scenario 2 is a physically unreasonable event while the probabilities of a significant loss of containment in scenarios 1 and 3 are extremely small. Before assessing the potential risk for the last two scenarios, the uncertainties in predicting complex phenomena for rare, high- consequence hazards needs to be addressed using a rigorous methodology.

  2. Hypothetical accident conditions thermal analysis of the 5320 package

    SciTech Connect (OSTI)

    Hensel, S.J.; Gromada, R.J.

    1995-12-31

    An axisymmetric model of the 5320 package was created to perform hypothetical accident conditions (HAC) thermal calculations. The analyses assume the 5320 package contains 359 grams of plutonium-238 (203 Watts) in the form of an oxide powder at a minimum density of 2.4 g/cc or at a maximum density of 11.2 g/cc. The solution from a non-solar 100 F ambient steady-state analysis was used as the initial conditions for the fire transient. A 30 minute 1,475 F fire transient followed by cooling via natural convection and thermal radiation to a 100 F non-solar environment was analyzed to determine peak component temperatures and vessel pressures. The 5320 package was considered to be horizontally suspended within the fire during the entire transient.

  3. Decontamination analysis of the NUWAX-83 accident site using DECON

    SciTech Connect (OSTI)

    Tawil, J.J.

    1983-11-01

    This report presents an analysis of the site restoration options for the NUWAX-83 site, at which an exercise was conducted involving a simulated nuclear weapons accident. This analysis was performed using a computer program deveoped by Pacific Northwest Laboratory. The computer program, called DECON, was designed to assist personnel engaged in the planning of decontamination activities. The many features of DECON that are used in this report demonstrate its potential usefulness as a site restoration planning tool. Strategies that are analyzed with DECON include: (1) employing a Quick-Vac option, under which selected surfaces are vacuumed before they can be rained on; (2) protecting surfaces against precipitation; (3) prohibiting specific operations on selected surfaces; (4) requiring specific methods to be used on selected surfaces; (5) evaluating the trade-off between cleanup standards and decontamination costs; and (6) varying of the cleanup standards according to expected exposure to surface.

  4. Technical evaluation: 300 Area steam line valve accident

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ``blanked off`` with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed.

  5. A Desk Guide to Buy American Provisions of the American Recovery and Reinvestment Act of 2009

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

    P a g e | 0 1 A Desk Guide to the Buy American Provisions of the Recovery Act- February 2011 2 A Desk Guide to the Buy American Provisions of the Recovery Act- February 2011 The US Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy (EERE) Buy American team has prepared this desk guide for Award Recipients of EERE Recovery Act Awards to assist them in applying the Buy American provisions 1 of the American Recovery and Reinvestment Act of 2009 (Recovery Act) to EERE awards

  6. Guidance on Incorporating EPA's Pollution Prevention Strategy...

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

    Incorporating EPA's Pollution Prevention Strategy into the Environmental Review Process Guidance on Incorporating EPA's Pollution Prevention Strategy into the Environmental Review ...

  7. Pollution Prevention - Environmental Impact Reduction Checklists...

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

    Pollution Prevention - Environmental Impact Reduction Checklists for NEPA309 Reviewers Pollution Prevention - Environmental Impact Reduction Checklists for NEPA309 Reviewers The ...

  8. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  9. Analysis of Kuosheng Station Blackout Accident Using MELCOR 1.8.4

    SciTech Connect (OSTI)

    Wang, S.-J.; Chien, C.-S.; Wang, T.-C.; Chiang, K.-S

    2000-11-15

    The MELCOR code, developed by Sandia National Laboratories, is a fully integrated, relatively fast-running code that models the progression of severe accidents in commercial light water nuclear power plants (NPPs).A specific station blackout (SBO) accident for Kuosheng (BWR-6) NPP is simulated using the MELCOR 1.8.4 code. The MELCOR input deck for Kuosheng NPP is established based on Kuosheng NPP design data and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The main severe accident phenomena and the fission product release fractions associated with the SBO accident were simulated. The predicted results are plausible and as expected in light of current understanding of severe accident phenomena. The uncertainty of this analysis is briefly discussed. The important features of the MELCOR 1.8.4 are described. The estimated results provide useful information for the probabilistic risk assessment (PRA) of Kuosheng NPP. This tool will be applied to the PRA, the severe accident analysis, and the severe accident management study of Kuosheng NPP in the near future.

  10. Analysis of Five Selected Tax Provisions of the Conference Energy Bill of 2003

    Reports and Publications (EIA)

    2004-01-01

    This special report was undertaken at the January 29, 2004, request of Senator John Sununu to perform an assessment of five specific tax provisions of the Conference Energy Bill of 2003.

  11. Attachment F: Summary Checklist of Model AIP Provisions and Program Policies

    Broader source: Energy.gov [DOE]

    In response to concerns expressed by DOE and State AIP program representatives, the Office of Environmental Activities has developed a list of mandatory and non-mandatory model language provisions...

  12. National Cost-Effectiveness Analysis of the Residential Provisions of the 2015 IECC

    SciTech Connect (OSTI)

    Mendon, Vrushali V.; Selvacanabady, Abinesh; Zhao, Mingjie; Taylor, Zachary T.

    2015-06-10

    This PNNL technical report summarizes the results of the national cost-effectiveness analysis of the residential provisions of the 2015 IECC compared to those of the 2012 and 2009 IECC.

  13. U-116: IBM Tivoli Provisioning Manager Express for Software Distribution Multiple Vulnerabilities

    Broader source: Energy.gov [DOE]

    Multiple vulnerabilities have been reported in IBM Tivoli Provisioning Manager Express for Software Distribution, which can be exploited by malicious people to conduct SQL injection attacks and compromise a user's system

  14. Colorado - C.R.S. 12-25-201 - Surveyors - General Provisions...

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Colorado - C.R.S. 12-25-201 - Surveyors - General ProvisionsLegal Abstract This section outlines the...

  15. Title 50 CFR 600 Magnuson-Stevens Act Provisions | Open Energy...

    Open Energy Info (EERE)

    50 CFR 600 Magnuson-Stevens Act Provisions Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- RegulationRegulation: Title 50 CFR 600...

  16. Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages

    SciTech Connect (OSTI)

    MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

    2003-01-01

    The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

  17. Summary Impacts of Modeled Provisions of the 2003 Conference Energy Bill

    Reports and Publications (EIA)

    2004-01-01

    This service report was undertaken at the February 2, 2004, request of Senator John Sununu to perform an assessment of the Conference Energy Bill of 2003. This report summarizes the CEB provisions that can be analyzed using the National Energy Modeling System (NEMS) and have the potential to affect energy consumption, supply, and prices. The impacts are estimated by comparing the projections with the CEB provisions to the AEO2004 Reference Case.

  18. The response of BWR Mark II containments to station blackout severe accident sequences

    SciTech Connect (OSTI)

    Greene, S.R.; Hodge, S.A.; Hyman, C.R.; Tobias, M.L. (Oak Ridge National Lab., TN (USA))

    1991-05-01

    This report describes the results of a series of calculations conducted to investigate the response of BWR Mark 2 containments to short-term and long-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to conduct quantitative accident sequence progression and containment response analyses for several station blackout scenarios. The accident mitigation effectiveness of automatic depressurization system actuation, drywell flooding via containment spray operation, and debris quenching in Mark 2 suppression pools is assessed. 27 refs., 16 figs., 21 tabs.

  19. Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident

    SciTech Connect (OSTI)

    Kun Zhang; Xuewu Cao [Shanghai Jiaotong University, Shanghai (China)

    2006-07-01

    The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG). (authors)

  20. Type A Accident Investigation Board Report on the April 3, 1995, Security

    Energy Savers [EERE]

    Rappel Tower Fatality at the DOE Savannah River Site | Department of Energy 3, 1995, Security Rappel Tower Fatality at the DOE Savannah River Site Type A Accident Investigation Board Report on the April 3, 1995, Security Rappel Tower Fatality at the DOE Savannah River Site August 1, 1995 The accident under investigation occurred on April 3, 1995, at approximately 10:46 a.m. As a result of the accident, a Wackenhut Services, Incorporated-Savannah River Site (WSI-SRS) Special Response Team

  1. Recent EPA pollution prevention initiatives

    SciTech Connect (OSTI)

    Bryant, C. )

    1993-01-01

    Today's rapidly developing and changing technologies and industrial practices frequently carry with them the increased generation of wastes and materials which, if improperly managed, may threaten public health and the environment. The US Environmental Protection Agency is charged with the mission of protecting public health and the environment from the hazards posed by these wastes and materials. As part of its effort to achieve this mandate, it has recently adopted a Pollution Prevention Program. Among other things, the program encourages the development and adoption of processing technologies and products that will lead to reducing the aggregate generation rates for pollutants entering the environment. This paper will address EPA's efforts under its Pollution Prevention Program. The paper will address regulatory and non-regulatory action EPA has taken pursuant to the Resource Conservation and Recovery Act (RCRA), the Emergency Planning and Community Right-to-Know Act, the Pollution Prevention Act of 1990, and other Federal statutes. In addition, the paper will present case studies in pollution prevention.

  2. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  3. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect (OSTI)

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  4. CASE STUDY FOR ENHANCED ACCIDENT TOLERANCE DESIGN CHANGES

    SciTech Connect (OSTI)

    Prescott, Steven; Smith, Curtis; Koonce, Tony

    2014-09-01

    The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies. The methods and tools provided by RISMC are essential to a comprehensive and integrated RIMM approach that supports effective preservation of margin for both active and passive SSCs. In this report, we discuss the methods and technologies behind RIMM for an application focused on enhanced accident tolerance design changes for a representative nuclear power plant. We look at a variety of potential plant modifications and evaluate, using the RISMC approach, the implications to safety margin for the various strategies.

  5. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  6. Level 1 Accident Report of the March 1, 2010 Bobcat Fatality...

    Office of Environmental Management (EM)

    March 31, 2010 On March 2, 2010 at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was convened to investigate an...

  7. A methodology for analyzing precursors to earthquake-initiated and fire-initiated accident sequences

    SciTech Connect (OSTI)

    Budnitz, R.J.; Lambert, H.E.; Apostolakis, G. and others

    1998-04-01

    This report covers work to develop a methodology for analyzing precursors to both earthquake-initiated and fire-initiated accidents at commercial nuclear power plants. Currently, the U.S. Nuclear Regulatory Commission sponsors a large ongoing project, the Accident Sequence Precursor project, to analyze the safety significance of other types of accident precursors, such as those arising from internally-initiated transients and pipe breaks, but earthquakes and fires are not within the current scope. The results of this project are that: (1) an overall step-by-step methodology has been developed for precursors to both fire-initiated and seismic-initiated potential accidents; (2) some stylized case-study examples are provided to demonstrate how the fully-developed methodology works in practice, and (3) a generic seismic-fragility date base for equipment is provided for use in seismic-precursors analyses. 44 refs., 23 figs., 16 tabs.

  8. The Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective

    Broader source: Energy.gov [DOE]

    The Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective Jeff Kimball Defense Nuclear Facilities Safety Board Staff Department of Energy NPH Conference October 26, 2011

  9. Overview of the U.S. DOE Accident Tolerant Fuel Development Program

    SciTech Connect (OSTI)

    Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton; Lance L. Snead

    2013-09-01

    The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.

  10. HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR LOSS-OF-COOLANT ACCIDENT

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

    (Technical Report) | SciTech Connect HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR LOSS-OF-COOLANT ACCIDENT Citation Details In-Document Search Title: HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR LOSS-OF-COOLANT ACCIDENT × You are accessing a document from the Department of Energy's (DOE) SciTech Connect. This site is a product of DOE's Office of Scientific and Technical Information (OSTI) and is provided as a public service. Visit OSTI to utilize additional information resources

  11. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect (OSTI)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  12. Type B Accident Investigation of the August 22, 2000, Injury Resulting From

    Office of Environmental Management (EM)

    Violent Exothermic Chemical Reaction at the Portsmouth Gaseous Diffusion Plant, X-701B Site | Department of Energy August 22, 2000, Injury Resulting From Violent Exothermic Chemical Reaction at the Portsmouth Gaseous Diffusion Plant, X-701B Site Type B Accident Investigation of the August 22, 2000, Injury Resulting From Violent Exothermic Chemical Reaction at the Portsmouth Gaseous Diffusion Plant, X-701B Site October 20, 2000 On August 22, 2000, an accident occurred at the U. S. Department

  13. Reactor safety study. An assessment of accident risks in U. S. commercial

    Office of Scientific and Technical Information (OSTI)

    nuclear power plants. Executive summary: main report. [PWR and BWR] (Technical Report) | SciTech Connect Technical Report: Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR] Citation Details In-Document Search Title: Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR] × You are accessing a document from the

  14. Type B Accident Investigation Board Report for the January 11, 2006,

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

    Personal Injury During Table Saw Use at the Heyrend Way Facility, Idaho Falls, Idaho | Department of Energy for the January 11, 2006, Personal Injury During Table Saw Use at the Heyrend Way Facility, Idaho Falls, Idaho Type B Accident Investigation Board Report for the January 11, 2006, Personal Injury During Table Saw Use at the Heyrend Way Facility, Idaho Falls, Idaho February 10, 2006 An accident at the Idaho National Laboratory (INL) was investigated in which a technician sustained a

  15. Type B Accident Investigation Board Report of the Savannah River Site Hand

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

    Injury at the Salt Waste Processing Facility on October 6, 2009 | Department of Energy Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009 Type B Accident Investigation Board Report of the Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009 November 1, 2009 This report documents the results of the Type B Accident Investigation Board (Board) investigation of the October 6, 2009, hand injury at the Department of Energy

  16. Type B Accident Investigation Report of the October 28, 2004, Burn Injuries

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

    Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA | Department of Energy of the October 28, 2004, Burn Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA Type B Accident Investigation Report of the October 28, 2004, Burn Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA February 1, 2005 TYPE B Accident Investigation

  17. Creep behavior of a nuclear pressure vessel under severe accident conditions

    SciTech Connect (OSTI)

    Beghini, M.; Bertini, L.; Vitale, E.

    1996-12-31

    The results of a study on the creep behavior of the vessel lower head under severe accident conditions are reported. An experimental program aimed at the evaluation of the creep properties of A533grB steel at high temperature (800--1,100 C) and under biaxial loading is summarized and the main results reported. A Finite Element simulation of the lower head under severe accident conditions allows to show the effect of the main parameters affecting the time to rupture.

  18. Thermal-stress analysis of a Fort St. Vrain core-support block under accident conditions

    SciTech Connect (OSTI)

    Carruthers, L.M.; Butler, T.A.; Anderson, C.A.

    1982-01-01

    A thermoelastic stress analysis of a graphite core support block in the Fort St. Vrain High Temperature Gas Cooled Reactor is described. The support block is subjected to thermal stresses caused by a loss of forced circulation accident of the reactor system. Two- and three-dimensional finite element models of the core support block are analyzed using the ADINAT and ADINA codes, and results are given that verify the integrity of this structural component under the given accident condition.

  19. Site restoration: Estimation of attributable costs from plutonium-dispersal accidents

    SciTech Connect (OSTI)

    Chanin, D.I.; Murfin, W.B.

    1996-05-01

    A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.

  20. Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

    SciTech Connect (OSTI)

    Putman, V.L.

    1995-09-01

    The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies.

  1. Analysis of Loss-of-Coolant Accidents in the NBSR

    SciTech Connect (OSTI)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  2. Environmental remediation following the Fukushima-Daiichi accident

    SciTech Connect (OSTI)

    Tagawa, A.; Miyahara, K.; Nakayama, S.

    2013-07-01

    A wide area of Fukushima Prefecture was contaminated with radioactivity released by the Fukushima Daiichi nuclear accident. The decontamination pilot projects conducted by JAEA aimed at demonstrating the applicability of different techniques to rehabilitate affected areas. As most radioactive cesium is concentrated at the top of the soil column and strongly bound to mineral surfaces, there are 3 options left to decrease the gamma dose rate (usually measured 1 m above the ground surface): the stripping of the contaminated topsoil (i.e. direct removal of cesium), the dilution by mixing and the soil profile inversion. The last two options do not generate waste. As the half-distance of {sup 137}Cs gammas in soil is in the order of 5-6 cm (depending on density and water content), the shielding by 50 cm of uncontaminated deep soil would theoretically reduce gamma doses by about 3 orders of magnitude. Which option is employed depends basically on the Cesium concentration in the topsoil, averaged over a 15-cm thickness. The JAEA's decontamination pilot projects focus on soil profile inversion and topsoil stripping. Two different techniques have been tested for the soil profile inversion: one is the reversal tillage by which surface soil of thickness of several tens of cm is reversed by using a tractor plough and the other is the complete interchanging of contaminated topsoil with uncontaminated subsoil by using a back-hoe. Reversal tillage with a tractor plough cost about 30 yen/m{sup 2}, which is an order of magnitude lower than that of topsoil-subsoil interchange (about 300 yen/m{sup 2}). Topsoil stripping is significantly more costly (between 550 yen/m{sup 2} and 690 yen/m{sup 2} according to the equipment used)

  3. Type B Accident Investigation of the Subcontractor Employee Injuries from a November 15, 2000, Fall Accident at the Oak Ridge National Laboratory

    Broader source: Energy.gov [DOE]

    On November 15, 2000, an accident occurred at the U. S. Department of Energy (DOE) Oak Ridge National Laboratory located in Oak Ridge, Tennessee. An employee of Decon and Recovery Services of Oak Ridge, LLC (DRS), working on an Oak Ridge Operations Office (ORO) Environmental Management decommissioning and demolition project received serious injuries from a fall (approximately 13 feet) from a fixed ladder.

  4. Energy Independence and Security Act of 2007: Summary of Provisions (released in AEO2008)

    Reports and Publications (EIA)

    2008-01-01

    The Energy Independence and Security Act of 2007 was signed into law on December 19, 2007, and became Public Law 110-140. Provisions in EISA2007 that require funding appropriations to be implemented, whose impact is highly uncertain, or that require further specification by federal agencies or Congress are not included in Annual Energy Outlook 2008 (AEO). For example, the Energy Information Administration (EIA) does not try to anticipate policy responses to the many studies required by EISA2007, nor to predict the impact of research and development (R&D) funding authorizations included in the bill. Moreover, AEO2008 does not include any provision that addresses a level of detail beyond that modeled in the National Energy Modeling System (NEMS), which was used to develop the AEO2008 projections. AEO2008 addresses only those provisions in EISA2007 that establish specific tax credits, incentives, or standards.

  5. FUEL CONSERVATION BY THE APPLICATION OF SPILL PREVENTION AND FAILSAFE ENGINEERING (A GUIDELINE MANUAL)

    SciTech Connect (OSTI)

    Goodier,, J. L.; Siclari,, R. J.; Garrity,, P. A.

    1980-10-30

    From a series of nationwide plant surveys dedicated to spill prevention, containment and countermeasure evaluation, coupled with spill response action activities, a need was determined for a spill prevention guideline manual. From Federally accumulated statistics for oil and hazardous substance spills, the authors culled information on spills of hydrocarbon products. In 1978, a total of 1456 oil spills were reported compared to 1451 in 1979. The 1978 spills were more severe, however, since 7;289,163 gallons of oil were accident~y discharged. In 1979, the gallons spilled was reduced to 3,663,473. These figures are derived from reported spills; it is highly possible that an equal amount was spilled and not reported. Spills effectively contained within a plant property that do not enter a n~vigational waterway need not be reported. Needless to say, there is a tremendous annual loss of oil products due to accidental spillage during transportation, cargo transfer, bulk storage and processing. As an aid to plant engineers and managers, Fe~eral workers, fire marshalls and fire and casualty insurance inspectors, the documen~ is offered as a spill prevention guide. The'manual defines state-of-the-art spill prevention practices and automation techniques that can reduce spills caused by human error. Whenever practical, the cost of implementation is provided to aid equipment acquisition and installation budgeting. To emphasize the need for spill prevention activities, historic spills are briefly described after which remedial action is defined in an appropriate section of the manual. The section on plant security goes into considerable depth since to date no Federal agency or traqe association has provided industry with guidelines on this important phase of plant operation. The intent of the document is to provide finger-tip reference material that can be used by interested parties in a nationwide effort to reduce loss of oil from preventable spills.

  6. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Hoover, M.D.; Newton, G.J.; Farrell, R.F.

    1996-06-01

    This qualitative hazard evaluation systematically assessed potential doses to workers during postulated accident conditions at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). Postulated accidents included the spontaneous ignition of a waste drum, puncture of a waste drum by a forklift, dropping of a waste drum from a forklift, and simultaneous dropping of seven drums during a crane failure. The descriptions and estimated frequencies of occurrence for these accidents were developed by the Hazard and Operability Study for CH TRU Waste Handling System (WCAP 14312). The estimated materials at risk, damage ratios, airborne release fractions and respirable fractions for these accidents were taken from the 1995 Safety Analysis Report (SAR) update and from the DOE handbook Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities (DOE-HDBK-3010-94). A Monte Carlo simulation was used to estimate the range of worker exposures that could result from each accident. Guidelines for evaluating the adequacy of defense-in-depth for worker protection at WIPP were adopted from a scheme presented by the International Commission on Radiological Protection in its publication on Protection from Potential Exposure: A Conceptual Framework (ICRP Publication 64). Probabilities of exposures greater than 5, 50, and 300 rem were less than 10{sup -2}, 10{sup -4}, and 10{sup -6} per year, respectively. In conformance with the guidance of DOE standard 3009-94, Appendix A (draft), we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposure under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, as well as members of the public and the environment.

  7. Sandia National Laboratories: Pollution Prevention: Regulatory...

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

    buildings and procurement contracts that integrate sustainability into their operations. ... Pollution Prevention Act Department of Energy Strategic Sustainability Performance Plan ...

  8. Preventing Proliferation and Developing Green Technologies |...

    National Nuclear Security Administration (NNSA)

    Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

  9. Preventive Maintenance Strategies for Compressed Air Systems...

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

    ... renewable energy, and pollution prevention technologies for industrial applications. ... paying attention to changes and trends, and responding promptly to maintain ...

  10. Preventive Maintenance Strategies for Compressed Air Systems

    Broader source: Energy.gov [DOE]

    This tip sheet discusses preventive maintenance strategies for compressed air systems to avoid high equipment repair and replacement costs.

  11. Pollution Prevention - Environmental Impact Reduction Checklists for

    Energy Savers [EERE]

    NEPA/309 Reviewers | Department of Energy Pollution Prevention - Environmental Impact Reduction Checklists for NEPA/309 Reviewers Pollution Prevention - Environmental Impact Reduction Checklists for NEPA/309 Reviewers The environmental review process under the National Environmental Policy Act (NEPA) provides a valuable opportunity for Federal agency NEPA/309 reviewers to incorporate pollution prevention and environmental impact reduction into actions (or projects). This Environmental

  12. ENHANCED SEVERE TRANSIENT ANALYSIS FOR PREVENTION TECHNICAL PROGRAM PLAN

    SciTech Connect (OSTI)

    Gougar, Hans

    2014-09-01

    This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.

  13. Pollution prevention cost savings potential

    SciTech Connect (OSTI)

    Celeste, J.

    1994-12-01

    The waste generated by DOE facilities is a serious problem that significantly impacts current operations, increases future waste management costs, and creates future environmental liabilities. Pollution Prevention (P2) emphasizes source reduction through improved manufacturing and process control technologies. This concept must be incorporated into DOE`s overall operating philosophy and should be an integral part of Total Quality Management (TQM) program. P2 reduces the amount of waste generated, the cost of environmental compliance and future liabilities, waste treatment, and transportation and disposal costs. To be effective, P2 must contribute to the bottom fine in reducing the cost of work performed. P2 activities at LLNL include: researching and developing innovative manufacturing; evaluating new technologies, products, and chemistries; using alternative cleaning and sensor technologies; performing Pollution Prevention Opportunity Assessments (PPOAs); and developing outreach programs with small business. Examples of industrial outreach are: innovative electroplating operations, printed circuit board manufacturing, and painting operations. LLNL can provide the infrastructure and technical expertise to address a wide variety of industrial concerns.

  14. Type B Accident Investigation Board Report on the May 24, 1998, Electrical Arc Blast at the Kansas City Plant

    Broader source: Energy.gov [DOE]

    This report is a product of an accident investigation board appointed by Bruce G. Twining, Manager, Albuquerque Operations Office, Department of Energy.

  15. Heat up and potential failure of BWR upper internals during a severe accident

    SciTech Connect (OSTI)

    Robb, Kevin R

    2015-01-01

    In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

  16. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect (OSTI)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  17. Environmental evaluation of municipal waste prevention

    SciTech Connect (OSTI)

    Gentil, Emmanuel C.; Gallo, Daniele; Christensen, Thomas H.

    2011-12-15

    Highlights: > Influence of prevention on waste management systems, excluding avoided production, is relatively minor. > Influence of prevention on overall supply chain, including avoided production is very significant. > Higher relative benefits of prevention are observed in waste management systems relying mainly on landfills. - Abstract: Waste prevention has been addressed in the literature in terms of the social and behavioural aspects, but very little quantitative assessment exists of the environmental benefits. Our study evaluates the environmental consequences of waste prevention on waste management systems and on the wider society, using life-cycle thinking. The partial prevention of unsolicited mail, beverage packaging and food waste is tested for a 'High-tech' waste management system relying on high energy and material recovery and for a 'Low-tech' waste management system with less recycling and relying on landfilling. Prevention of 13% of the waste mass entering the waste management system generates a reduction of loads and savings in the waste management system for the different impacts categories; 45% net reduction for nutrient enrichment and 12% reduction for global warming potential. When expanding our system and including avoided production incurred by the prevention measures, large savings are observed (15-fold improvement for nutrient enrichment and 2-fold for global warming potential). Prevention of food waste has the highest environmental impact saving. Prevention generates relatively higher overall relative benefit for 'Low-tech' systems depending on landfilling. The paper provides clear evidence of the environmental benefits of waste prevention and has specific relevance in climate change mitigation.

  18. Analysis of Selected Provisions of the Domestic Manufacturing and Energy Jobs Act of 2010

    Reports and Publications (EIA)

    2010-01-01

    This report responds to a letter dated August 16, 2010, from Janice Mays, Staff Director of the U.S. House of Representatives' Committee on Ways and Means, requesting that the U.S. Energy Information Administration (EIA) analyze several provisions included in the July 26, 2010, discussion draft of the Domestic Manufacturing and Energy Jobs Act of 2010.

  19. Energy Improvement and Extension Act of 2008: Summary of Provisions (released in AEO2009)

    Reports and Publications (EIA)

    2009-01-01

    The Emergency Economic Stabilization Act of 2008 (Public Law 110-343), which was signed into law on October 3, 2008, incorporates EIEA2008 in Division B. Provisions in EIEA2008 that require funding appropriations to be implemented, whose impact is highly uncertain or that require further specification by federal agencies or Congress, are not included in Annual Energy Outlook 2009.

  20. Trends in state-level freight accident rates: An enhancement of risk factor development for RADTRAN

    SciTech Connect (OSTI)

    Saricks, C.; Kvitek, T.

    1991-01-01

    Under the Nuclear Waste Policy Act, the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) is concerned with understanding and managing risk as it applies to the shipment of spent commercial nuclear reactor fuel. Understanding risk in relation to mode and geography may provide opportunities to minimize radiological and non-radiological risks of transportation. To enhance such an understanding, a set of state-or waterway-specific accident, fatality, and injury rates (expressed as rates per shipment kilometer) by transportation mode and highway administrative class was developed, using publicly-available data bases. Adjustments made to accommodate miscoded or incomplete information in accident data are described, as well as the procedures for estimating state-level flow data. Results indicate that the shipping conditions under which spent fuel is likely to be transported should be less subject to accidents than the average'' shipment within mode. 10 refs., 3 tabs.

  1. Final safety analysis report for the Galileo Mission: Volume 2, Book 2: Accident model document: Appendices

    SciTech Connect (OSTI)

    Not Available

    1988-12-15

    This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

  2. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect (OSTI)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  3. Accident Generated Particulate Materials and Their Characteristics -- A Review of Background Information

    SciTech Connect (OSTI)

    Sutter, S. L.

    1982-05-01

    Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of the amount of radioactive particulate material initially airborne (source term) during accidents. Pacific Northwest Laboratory (PNL) has surveyed the literature, gathering information on the amount and size of these particles that has been developed from limited experimental work, measurements made from operational accidents, and known aerosol behavior. Information useful for calculating both liquid and powder source terms is compiled in this report. Potential aerosol generating events discussed are spills, resuspension, aerodynamic entrainment, explosions and pressurized releases, comminution, and airborne chemical reactions. A discussion of liquid behavior in sprays, sparging, evaporation, and condensation as applied to accident situations is also included.

  4. Superheated-steam test of ethylene propylene rubber cables using a simultaneous aging and accident environment

    SciTech Connect (OSTI)

    Bennett, P.R.; St. Clair, S.D.; Gilmore, T.W.

    1986-06-01

    The superheated-steam test exposed different ethylene propylene rubber (EPR) cables and insulation specimens to simultaneous aging and a 21-day simultaneous accident environment. In addition, some insulation specimens were exposed to five different aging conditions prior to the 21-day simultaneous accident simulation. The purpose of this superheated-steam test (a follow-on to the saturated-steam tests (NUREG/CR-3538)) was to: (1) examine electrical degradation of different configurations of EPR cables; (2) investigate differences between using superheated-steam or saturated-steam at the start of an accident simulation; (3) determine whether the aging technique used in the saturated-steam test induced artificial degradation; and (4) identify the constituents in EPR that affect moisture absorption.

  5. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  6. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  7. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect (OSTI)

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

  8. Development of LWR Fuels with Enhanced Accident Tolerance

    SciTech Connect (OSTI)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.

  9. NNSA Releases New Nuclear Prevent, Counter, and Respond Report...

    National Nuclear Security Administration (NNSA)

    People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy ... on NNSA's efforts to prevent nuclear proliferation and terrorism, Prevent, Counter, and ...

  10. Technical Advisory Team (TAT) report on the rocket sled test accident of October 9, 2008.

    SciTech Connect (OSTI)

    Stofleth, Jerome H.; Dinallo, Michael Anthony; Medina, Anthony J.

    2009-01-01

    This report summarizes probable causes and contributing factors that led to a rocket motor initiating prematurely while employees were preparing instrumentation for an AIII rocket sled test at SNL/NM, resulting in a Type-B Accident. Originally prepared by the Technical Advisory Team that provided technical assistance to the NNSA's Accident Investigation Board, the report includes analyses of several proposed causes and concludes that the most probable source of power for premature initiation of the rocket motor was the independent battery contained in the HiCap recorder package. The report includes data, evidence, and proposed scenarios to substantiate the analyses.

  11. The February 2014 Accidents at WIPP - What Happened and What We Know About Why

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

    March 19, 2015, Phoenix, Arizona, USA 1 The February 2014 Accidents at WIPP - 15024 (What Happened and What We Know About Why) Roger Nelson, Russel Patterson, Abe VanLuik U.S. Department of Energy, PO Box 3090, Carlsbad, NM 88220 (roger.nelson@wipp.ws) ABSTRACT With almost 15 years of successful and safe operations, the WIPP facility was suddenly shutdown in February 2014 due to two unrelated accidents underground. A fire burned the front tires and engine of a salt haul truck, creating

  12. Y-12 Construction hits one million-hour mark without a lost-time accident |

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

    Y-12 National Security Complex Construction hits one ... Y-12 Construction hits one million-hour mark without a lost-time accident Posted: August 30, 2012 - 5:30pm The B&W Y-12 Direct-Hire Construction team has worked one million hours, covering a 633-day period, without a lost-time injury. Some 285 people including building trade crafts, non-manual staff and escorts worked without a lost-time accident during this period. The Construction team's last lost workday was in September 2010. A

  13. Accident Investigation of the July 30, 2013, Electrical Fatality on the

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

    Bandon-Rogue No. 1 115kV Line at the Bonneville Power Administration | Department of Energy July 30, 2013, Electrical Fatality on the Bandon-Rogue No. 1 115kV Line at the Bonneville Power Administration Accident Investigation of the July 30, 2013, Electrical Fatality on the Bandon-Rogue No. 1 115kV Line at the Bonneville Power Administration July 30, 2013 On August 7, 2013, at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was

  14. Fatal accidents involving roof falls in coal mining, 1996--1998

    SciTech Connect (OSTI)

    Not Available

    1999-01-01

    This publication presents information on fatalities involving roof and rib falls that occurred in coal mining operations from January 1996 through December 1998. It includes statistics for the fatalities, as well as abstracts, best practices and illustrations. Conclusion statements have been substituted for best practices where no Title 30 Code of Regulations violations were cited during the accident investigation. From January 1996 through December 1998, 36 miners died at coal operations from accidents classified as roof falls. The information in the report is based on statistics taken from the 1996 through 1998 MSHA Fatal Illustration Programs: Roof Fall Fatalities by District.

  15. Fatal accidents involving roof falls in coal mining, 1996--1998

    SciTech Connect (OSTI)

    1999-11-01

    This publication presents information on fatalities involving roof and rib falls that occurred in coal mining operations from January 1996 through December 1998. It includes statistics for the fatalities, as well as abstracts, best practices and illustrations. Conclusion statements have been substituted for best practices where no Title 30 Code of Regulations violations were cited during the accident investigation. From January 1996 through December 1998, 36 miners died at coal operations from accidents classified as roof falls. The information in the report is based on statistics taken from the 1996 through 1998 MSHA Fatal Illustration Programs: Roof Fall Fatalities by District.

  16. Type B Accident Investigation Board Report BNFL, Inc. Employee Foot Injury

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

    on December 17, 2003, at the East Tennessee Technology Park Building K-31 | Department of Energy BNFL, Inc. Employee Foot Injury on December 17, 2003, at the East Tennessee Technology Park Building K-31 Type B Accident Investigation Board Report BNFL, Inc. Employee Foot Injury on December 17, 2003, at the East Tennessee Technology Park Building K-31 February 1, 2004 On December 17, 2003, at approximately 7:15 a.m., an accident occurred at the U.S. Department of Energy (DOE) East Tennessee

  17. Type B Accident Investigation Board Report Employee Puncture Wound at the

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

    F-TRU Waste Remediation Facility at the Savannah River Site on June 14, 2010 | Department of Energy Employee Puncture Wound at the F-TRU Waste Remediation Facility at the Savannah River Site on June 14, 2010 Type B Accident Investigation Board Report Employee Puncture Wound at the F-TRU Waste Remediation Facility at the Savannah River Site on June 14, 2010 September 1, 2010 This report documents the results of the Type B Accident Investigation Board investigation of the June 14, 2010,

  18. Type B Accident Investigation of the Savannah River Site Arc Flash Burn

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

    Injury on September 23, 2009, in the D Area Powerhouse | Department of Energy of the Savannah River Site Arc Flash Burn Injury on September 23, 2009, in the D Area Powerhouse Type B Accident Investigation of the Savannah River Site Arc Flash Burn Injury on September 23, 2009, in the D Area Powerhouse October 1, 2009 This report documents the results of the Type B Accident Investigation Board investigation of the September 23, 2009, employee burn injury at the Department of Energy (DOE)

  19. Improvement of Design Codes to Account for Accident Thermal Effects on Seismic Performance

    Office of Environmental Management (EM)

    IMPROVEMENT OF DESIGN CODES TO ACCOUNT FOR ACCIDENT THERMAL EFFECTS ON SEISMIC PERFORMANCE Amit H. Varma, Kadir Sener, Saahas Bhardwaj Purdue University Andrew Whittaker: Univ. of Buffalo INTRODUCTION  Project focuses on the effects of accident thermal conditions on the seismic performance of: a) Innovative steel-plate composite SC walls, and b) Conventional reinforced concrete RC walls. S t e e l F a c e p l a t e s P e n e t r a t i o n A t t a c h m e n t C o n c r e t e T i e B a r s

  20. Applications of human factors engineering to LNG release prevention and control

    SciTech Connect (OSTI)

    Shikiar, R.; Rankin, W.L.; Rideout, T.B.

    1982-06-01

    The results of an investigation of human factors engineering and human reliability applications to LNG release prevention and control are reported. The report includes a discussion of possible human error contributions to previous LNG accidents and incidents, and a discussion of generic HF considerations for peakshaving plants. More specific recommendations for improving HF practices at peakshaving plants are offered based on visits to six facilities. The HF aspects of the recently promulgated DOT regulations are reviewed, and recommendations are made concerning how these regulations can be implemented utilizing standard HF practices. Finally, the integration of HF considerations into overall system safety is illustrated by a presentation of human error probabilities applicable to LNG operations and by an expanded fault tree analysis which explicitly recognizes man-machine interfaces.

  1. Los Alamos wins 2008 Pollution Prevention awards

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

    2008 Pollution Prevention awards Los Alamos wins 2008 Pollution Prevention awards Winner of two Best-in-Class Pollution Prevention awards and six Environmental Stewardship awards from the National Nuclear Security Administration. February 7, 2008 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new

  2. Site Discharge Pollution Prevention Plan (SDPPP)

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

    SDPPP Individual Permit: Site Discharge Pollution Prevention Plan (SDPPP) The 2014 SDPPP update fully incorporates all changes made during the year and reflects changes projected...

  3. Jefferson Lab Stormwater Pollution Prevention Reminder | Jefferson...

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

    Stormwater Pollution Prevention Reminder Stormwater runoff occurs when rainfall or snowmelt flows over ground surfaces. Naturally vegetated ground surfaces often absorb the...

  4. Calculation notes that support accident scenario and consequence development for the leak from a railcar/tank trailer at the 204-ar waste unloading facility

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Leak from Railcar/Tank Trailer. The calculations needed to quantify the risk associated with this accident scenario are included within.

  5. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  6. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    SciTech Connect (OSTI)

    2013-09-25

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  7. Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

    SciTech Connect (OSTI)

    Rempe, Joy L.; Knudson, Darrell L.

    2015-02-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

  8. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema (OSTI)

    None

    2014-03-11

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  9. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M.

    2014-03-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilities while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.

  10. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilitiesmore »while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.« less

  11. Grandfathering provisions

    U.S. Energy Information Administration (EIA) Indexed Site

    Key Changes Required by EISA Energy Independence and Security Act (December 2007) required changes to the RFS program Significantly increased volumes of renewable fuel - to ...

  12. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect (OSTI)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

  13. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    SciTech Connect (OSTI)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk.

  14. 10 C.F.R. PART 835-OCCUPATIONAL RADIATION PROTECTION, Subpart A - General Provisions

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

    835-OCCUPATIONAL RADIATION PROTECTION Authority: 42 U.S.C. 2201; 7191. Source: 58 FR 65485, Dec. 14, 1993, unless otherwise noted. Subpart A-General Provisions § 835.1 Scope. (a) General. The rules in this part establish radiation protection standards, limits, and program requirements for protecting individuals from ionizing radiation resulting from the conduct of DOE activities. (b) Exclusion. The requirements in this part do not apply to: 1. Activities that are regulated through a license by

  15. American Recovery and Reinvestment Act of 2009: Summary of Provisions (released in AEO2010)

    Reports and Publications (EIA)

    2010-01-01

    The American Recovery and Reinvestment Act of 2009 (ARRA), signed into law in mid-February 2009, provides significant new federal funding, loan guarantees, and tax credits to stimulate investments in energy efficiency and renewable energy. The provisions of ARRA were incorporated initially as part of a revision to the Annual Energy Outlook 2009 Reference case that was released in April 2009, and they also are included in Annual Energy Outlook 2010.

  16. Isolation and Performance Guarantee Provisioning in Argo NodeOS with Custom

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

    Compute Containers | Argonne Leadership Computing Facility Isolation and Performance Guarantee Provisioning in Argo NodeOS with Custom Compute Containers Event Sponsor: Mathematics and Computer Science Division Seminar Start Date: Feb 23 2016 - 12:00pm Building/Room: Building 240/Room 4301 Location: Argonne National Laboratory Speaker(s): Judicael Zounmevo Host: Kate Keahey We will discuss the Argo NodeOS single-kernel approach to OS specialization via our in-development compute

  17. Final Rule on Amending Eligibility Provisions to Multifamily Buildings for the Weatherization Assistance Program

    Energy Savers [EERE]

    WEATHERIZATION PROGRAM NOTICE 10-15 EFFECTIVE DATE: March 2, 2010 SUBJECT: FINAL RULE ON AMENDING ELIGIBILITY PROVISIONS TO MULTI-FAMILY BUILDINGS FOR THE WEATHERIZATION ASSISTANCE PROGRAM PURPOSE: To issue guidance on implementing recent changes to the Weatherization Assistance Program (WAP) requirements for determining eligibility of certain multi-family buildings as identified by the Department of Housing and Urban Development (HUD) and the Department of Agriculture (USDA). LEGAL AUTHORITY:

  18. Meeting pollution prevention goals: Successful implementation

    SciTech Connect (OSTI)

    Seith, B.J. )

    1993-01-01

    This paper focuses on the essential, but often overlooked, elements of a pollution prevention program: the steps required for a successful implementation. As programs are being developed, attention must be given to assuring that the systems to support a successful introduction and continued improvement are in place. The goals of a pollution prevention plan (i.e. 50% reduction in toxics use and 40% reduction in hazardous waste generation within three years) must be translated into performance oriented-responsibilities taken throughout an organization, at all levels. Successful implementation requires a genuine commitment from management, employee awareness programs tailored to each type of audience, and a feedback system to assure that the program is continually changing to incorporate new pollution prevention challenges. Also, by conducting an economic analysis of pollution prevention opportunities and activities, and incorporating the results into the business decision-making process, a company is more apt to make wise and measurable performance towards its pollution prevention goals.

  19. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO? fuel and APMT-Steel Clad/UN-U?Si? fuel concepts

    SciTech Connect (OSTI)

    Unal, Cetin; Galloway, Jack D.

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO? ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U?Si? fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  20. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect (OSTI)

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  1. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

    SciTech Connect (OSTI)

    Himes, D.A.

    1997-11-17

    The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines.

  2. Source terms released into the environment for a station blackout severe accident at the Peach Bottom Atomic Power Station

    SciTech Connect (OSTI)

    Carbajo, J.J.

    1995-07-01

    This study calculates source terms released into the environment at the Peach Bottom Atomic Power Station after containment failure during a postulated low-pressure, short-term station blackout severe accident. The severe accident analysis code MELCOR, version 1.8.1, was used in these calculations. Source terms were calculated for three different containment failure modes. The largest environmental releases occur for early containment failure at the drywell liner in contact with the cavity by liner melt-through. This containment failure mode is very likely to occur when the cavity is dry during this postulated severe accident sequence.

  3. Preventing Space Traffic Jams | Department of Energy

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

    Preventing Space Traffic Jams Preventing Space Traffic Jams June 10, 2015 - 2:25pm Addthis With around one thousand active satellites and the tens of thousands of pieces of space junk orbiting Earth, space is getting exceedingly crowded. Researchers at Lawrence Livermore National Lab are working on a system that could help prevent collisions in space. | Graphic by <a href="/node/379579">Sarah Gerrity</a>, Energy Department. With around one thousand active satellites and the

  4. Type A Accident Investigation Board Report on the January 17, 1996, Electrical Accident With Injury in Building 209, Technical Area 21, Tritium Science and Fabrication Facility, Los Alamos National Laboratory

    Office of Energy Efficiency and Renewable Energy (EERE)

    This report is an independent product of the Type A Accident Investigation Board appointed by Tara O’Toole, M.D., M.P.H., Assistant Secretary for Environment, Safety and Health (EH-1).

  5. Accident Investigation at the Idaho National Laboratory Engineering Demonstration Facility, February 2013

    Broader source: Energy.gov [DOE]

    On Monday, February 12, 2013, a principal investigator at the Idaho National Laboratory (INL) Engineering Demonstration Facility (IEDF) was testing the system configuration of experimental process involving liquid sodium carbonate. An unanticipated event occurred that resulted in the ejection of the 900° C liquid sodium carbonate from the system. The ejected liquid came into contact with the principal investigator and caused multiple second and third degree burn injuries to approximately 10 percent of his body. The Office of Health, Safety and Security (HSS) Site Lead for the Idaho Site shadowed the accident investigation team assembled by the contractor in an effort to independently verify that a rigorous, thorough, and unbiased investigation was taking place, and to maintain awareness of the events surrounding the accident

  6. Introduction to the Special Issue on the U.S. Response to the Fukushima Accident

    SciTech Connect (OSTI)

    Blumenthal, Daniel J.

    2012-05-01

    Provides an introduction to the May 2012 issue of Health Physics, based on a special session at the 2011 Health Physics Society (HPS) annual meeting that focused on the United States' radiological response to the Fukushima Daiichi Nuclear Power Plant accident. This introduction outlines the papers in this important issue and describes the activities of the U.S. response participants, including the U.S. Department of Energy National Nuclear Security Administration (DOE/NNSA), Department of Defense, the U.S. Nuclear Regulatory Commission (NRC) and other organizations. Observations are provided and the stage is set for the articles in this issue which document many of the activities undertaken during the Fukushima accident and which describe challenges faced and valuable lessons learned.

  7. Aerosol Sample Inhomogeneity with Debris from the Fukushima Daiichi Nuclear Accident

    SciTech Connect (OSTI)

    Gomez, Reynaido; Biegalski, Steven R.; Woods, Vincent T.

    2014-09-01

    Radionuclide aerosol sampling is a vital component in the detection of nuclear explosions, nuclear accidents, and other radiation releases. This was proven by the detection and tracking of emissions from the Fukushima Daiichi incident across the globe by IMS stations. Two separate aerosol samplers were operated in Richland, WA following the event and debris from the accident were measured at levels well above detection limits. While the atmospheric activity concentration of radionuclides generally compared well between the two stations, they did not agree within uncertainties. This paper includes a detailed study of the aerosol sample homogeneity of 134Cs and 137Cs, then relates it to the overall uncertainty of the original measurement. Our results show that sample inhomogeneity adds an additional 5–10% uncertainty to each aerosol measurement and that this uncertainty is in the same range as the discrepancies between the two aerosol sample measurements from Richland, WA.

  8. Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents. [HTGR

    SciTech Connect (OSTI)

    Houghton, W.J.

    1980-06-01

    A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors (HTGRs).

  9. Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents.

    SciTech Connect (OSTI)

    Majumdar, S.; Diercks, D. R.; Shack, W. J.; Energy Technology

    2002-05-01

    This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents.

  10. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  11. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    SciTech Connect (OSTI)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  12. Bounding Radionuclide Inventory and Accident Consequence Calculation for the 1L Target

    SciTech Connect (OSTI)

    Kelsey, Charles T. IV

    2011-01-01

    A bounding radionuclide inventory for the tungsten of the Los Alamos Neutron Science Center (LANSCE) IL Target is calculated. Based on the bounding inventory, the dose resulting from the maximum credible incident (MCI) is calculated for the maximally exposed offsite individual (MEOl). The design basis accident involves tungsten target oxidation following a loss of cooling accident. Also calculated for the bounding radionuclide inventory is the ratio to the LANSCE inventory threshold for purposes of inventory control as described in the target inventory control policy. A bounding radionuclide inventory calculation for the lL Target was completed using the MCNPX and CINDER'90 codes. Continuous beam delivery at 200 {micro}A to 2500 mA{center_dot}h was assumed. The total calculated activity following this irradiation period is 205,000 Ci. The dose to the MEOI from the MCI is 213 mrem for the bounding inventory. The LANSCE inventory control threshold ratio is 132.

  13. Final report of the accident phenomenology and consequence (APAC) methodology evaluation. Spills Working Group

    SciTech Connect (OSTI)

    Brereton, S.; Shinn, J.; Hesse, D; Kaninich, D.; Lazaro, M.; Mubayi, V.

    1997-08-01

    The Spills Working Group was one of six working groups established under the Accident Phenomenology and Consequence (APAC) methodology evaluation program. The objectives of APAC were to assess methodologies available in the accident phenomenology and consequence analysis area and to evaluate their adequacy for use in preparing DOE facility safety basis documentation, such as Basis for Interim Operation (BIO), Justification for Continued Operation (JCO), Hazard Analysis Documents, and Safety Analysis Reports (SARs). Additional objectives of APAC were to identify development needs and to define standard practices to be followed in the analyses supporting facility safety basis documentation. The Spills Working Group focused on methodologies for estimating four types of spill source terms: liquid chemical spills and evaporation, pressurized liquid/gas releases, solid spills and resuspension/sublimation, and resuspension of particulate matter from liquid spills.

  14. A methodology for generating dynamic accident progression event trees for level-2 PRA

    SciTech Connect (OSTI)

    Hakobyan, A.; Denning, R.; Aldemir, T. [Ohio State Univ., Nuclear Engineering Program, 650 Ackerman Road, Columbus, OH 43202 (United States); Dunagan, S.; Kunsman, D. [Sandia National Laboratory, Albuquerque, NM 87185 (United States)

    2006-07-01

    Currently, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. A software tool (ADAPT) is described for automated APET generation using the concept of dynamic event trees. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. While the software tool could be applied to any systems analysis code, the MELCOR code is used for this illustration. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a pressurized water reactor. (authors)

  15. Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

    SciTech Connect (OSTI)

    Griffin, F.P.

    1995-12-31

    A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B{sub 4}C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence.

  16. Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

    2014-09-01

    The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

  17. Fukushima Daiichi Accident Study Information Portal Quality Assurance Review: Pre-Public Release

    SciTech Connect (OSTI)

    Kurt G. Vedros

    2012-01-01

    This design review compared the current product with the original intent set forth in the initial internet portal design found in the document: Fukushima Daiichi Database Design, Revision 5. The current revision of the Fukushima Daiichi Accident Study Information Portal (FDASIP) is 1.0.21. This revision is one that restricts access for each user based on roles granted by the project administrator. The public access revision is currently on the test server and will be considered in this review as well.

  18. The Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective

    Office of Environmental Management (EM)

    Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective Jeff Kimball Defense Nuclear Facilities Safety Board Staff Department of Energy NPH Conference October 26, 2011 The views expressed are solely those of the author and no official support or endorsement of this presentation by the Defense Nuclear Facilities Safety Board or the federal government is intended or should be inferred. 1 OBJECTIVE: Assess whether the DOE NPH performance goal concept as

  19. Accident Investigation of the September 20, 2012 Fatal Fall from the

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

    Dworshak-Taft #1 Transmission Tower, at the Bonneville Power Marketing Administration | Department of Energy September 20, 2012 Fatal Fall from the Dworshak-Taft #1 Transmission Tower, at the Bonneville Power Marketing Administration Accident Investigation of the September 20, 2012 Fatal Fall from the Dworshak-Taft #1 Transmission Tower, at the Bonneville Power Marketing Administration November 30, 2012 On September 21, 2012, at the request of the Bonneville Power Administration (BPA) Chief

  20. Advanced surface plasma nitriding for development of corrosion resistant and accident tolerant fuel cladding

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

    surface plasma nitriding for development of corrosion resistant and accident tolerant fuel cladding PI: Lin Shao, Texas A&M University Collaborators: Don A. Lucca - Oklahoma State University, Michael P. Short - Massachusetts Institute of Technology, Frank Garner - Texas A&M University Program: Nuclear Energy Enabling Technologies Presented by Elizabeth Castanon, Ph.D candidate in Nuclear Eng., Texas A&M University This project aims to develop a new plasma nitriding technique which is

  1. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    SciTech Connect (OSTI)

    Nourbakhsh, H.P.

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented.

  2. Effects of spent fuel types on offsite consequences of hypothetical accidents

    SciTech Connect (OSTI)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-02-18

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor ({chi}/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only {sup 134}Cs, {sup 137}Cs-{sup 137m}Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents.

  3. Synthesis of VERCORS and Phebus data in severe accident codes and applications.

    SciTech Connect (OSTI)

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.

  4. A probabilistic risk assessment of the LLNL Plutonium facility`s evaluation basis fire operational accident

    SciTech Connect (OSTI)

    Brumburgh, G.

    1994-08-31

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility.

  5. Type A Accident Investigation Board Report on the April 19, 1999, Special

    Energy Savers [EERE]

    Agent Fatality at the Southeast Courier Section, Oak Ridge, Tennessee | Department of Energy 19, 1999, Special Agent Fatality at the Southeast Courier Section, Oak Ridge, Tennessee Type A Accident Investigation Board Report on the April 19, 1999, Special Agent Fatality at the Southeast Courier Section, Oak Ridge, Tennessee June 1, 1999 On April 19, 1999, a fatality occurred at the Albuquerque Operations Office (AL) Southeast Courier Section (SCS). A Special Agent had finished his 1-mile

  6. Type A Accident Investigation of the June 21, 2001, Drilling Rig Operator

    Energy Savers [EERE]

    Injury at the Fermi National Accelerator Laboratory, August 2001 | Department of Energy June 21, 2001, Drilling Rig Operator Injury at the Fermi National Accelerator Laboratory, August 2001 Type A Accident Investigation of the June 21, 2001, Drilling Rig Operator Injury at the Fermi National Accelerator Laboratory, August 2001 August 8, 2001 On June 21, 2001, at approximately 9:40 A.M., a construction sub-tier contractor employee (the "Operator") at the Fermi National Accelerator

  7. Type A Accident Investigation of the March 16, 2000, Plutonium-238 Multiple

    Energy Savers [EERE]

    Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico | Department of Energy March 16, 2000, Plutonium-238 Multiple Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico Type A Accident Investigation of the March 16, 2000, Plutonium-238 Multiple Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico July 1, 2000 On March 16, 2000, at approximately 2 p.m., a radiological release of plutonium-238 occurred

  8. Type A Accident Report of the June 26, 2009 Vehicle Fatality at Lawrence

    Energy Savers [EERE]

    Livermore National Laboratory | Department of Energy Report of the June 26, 2009 Vehicle Fatality at Lawrence Livermore National Laboratory Type A Accident Report of the June 26, 2009 Vehicle Fatality at Lawrence Livermore National Laboratory October 1, 2009 On June 26, 2009, a Lawrence Livermore National Security (LLNS) employee was in the process of transporting six boxes containing personal property to his new office in preparation for a routine transfer to another position within the

  9. Type B Accident Investigation Board Report of the Bechtel Jacobs Company,

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

    LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee | Department of Energy Bechtel Jacobs Company, LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board Report of the Bechtel Jacobs Company, LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee February 10,

  10. Type B Accident Investigation Board Report of the September 29, 2010,

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

    Radiological Contamination Event at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna, New, York | Department of Energy September 29, 2010, Radiological Contamination Event at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna, New, York Type B Accident Investigation Board Report of the September 29, 2010, Radiological Contamination Event at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna,

  11. Type B Accident Investigation Board Report on the October 15, 2001, Grout

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

    Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental Laboratory | Department of Energy 15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental Laboratory Type B Accident Investigation Board Report on the October 15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental Laboratory November 13, 2001 On October 15, 2001, at

  12. Type B Accident Investigation Board Report on the October 8, 2004, Grounds

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

    Worker Injury at the Pacific Northwest National Laboratory | Department of Energy 8, 2004, Grounds Worker Injury at the Pacific Northwest National Laboratory Type B Accident Investigation Board Report on the October 8, 2004, Grounds Worker Injury at the Pacific Northwest National Laboratory November 22, 2004 On October 8, 2004, at approximately 11:00 am, a Grounds Worker at the Pacific Northwest National laboratory (PNNL) fell from a Toro Workman 3200 Utility Vehicle and fracturedhis right

  13. Type B Accident Investigation Board Report, May 8, 2004, Exothermic Metal

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

    Reactor Event During Sodium Transfer Activities, East Tennessee Technology Park, Oak Ridge, Tennessee | Department of Energy Report, May 8, 2004, Exothermic Metal Reactor Event During Sodium Transfer Activities, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board Report, May 8, 2004, Exothermic Metal Reactor Event During Sodium Transfer Activities, East Tennessee Technology Park, Oak Ridge, Tennessee August 17, 2004 On May 8, 2004, at approximately 11:00

  14. Type B Accident Investigation of the October 9, 2008 Employee Injured when

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

    Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office | Department of Energy of the October 9, 2008 Employee Injured when Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office Type B Accident Investigation of the October 9, 2008 Employee Injured when Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office

  15. Type B Accident Investigation on the June 27, 2002, Exothermic Metal

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

    Reaction Event During Converter Disassembly in Building K-33 at the East Tennessee Technology Park | Department of Energy on the June 27, 2002, Exothermic Metal Reaction Event During Converter Disassembly in Building K-33 at the East Tennessee Technology Park Type B Accident Investigation on the June 27, 2002, Exothermic Metal Reaction Event During Converter Disassembly in Building K-33 at the East Tennessee Technology Park July 31, 2002 On June 27, 2002, an exothermic metal reaction

  16. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    SciTech Connect (OSTI)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, water ingestion dose, milk growing season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meat, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of 1-131 at which disposal of crops will be initiated due to accidents that occur during the growing season.

  17. DOE-STD-3014-96; DOE Standard Accident Analysis For Aircraft Crash Into Hazardous Facilities

    Office of Environmental Management (EM)

    DOE STANDARD October 1 996 Reaffirmation May 2006 ACCIDENT CRASH INTO ANALYSIS HAZARDOUS FACILITIES U.S. Department of Energy Washington, DC 20585 AREA SAFT DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. FOR AIRCRAFT This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; (423) 576-8401. Available to the public from

  18. Guidance on Incorporating EPA's Pollution Prevention Strategy into the

    Office of Environmental Management (EM)

    Environmental Review Process | Department of Energy on Incorporating EPA's Pollution Prevention Strategy into the Environmental Review Process Guidance on Incorporating EPA's Pollution Prevention Strategy into the Environmental Review Process The guidance discusses the Environmental Protection Agency's definition of pollution prevention; how to incorporate pollution prevention into the EPA environmental review process and interagency liaison function; and federal pollution prevention awards

  19. Model development for household waste prevention behaviour

    SciTech Connect (OSTI)

    Bortoleto, Ana Paula; Kurisu, Kiyo H.; Hanaki, Keisuke

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We model waste prevention behaviour using structure equation modelling. Black-Right-Pointing-Pointer We merge attitude-behaviour theories with wider models from environmental psychology. Black-Right-Pointing-Pointer Personal norms and perceived behaviour control are the main behaviour predictors. Black-Right-Pointing-Pointer Environmental concern, moral obligation and inconvenience are the main influence on the behaviour. Black-Right-Pointing-Pointer Waste prevention and recycling are different dimensions of waste management behaviour. - Abstract: Understanding waste prevention behaviour (WPB) could enable local governments and decision makers to design more-effective policies for reducing the amount of waste that is generated. By merging well-known attitude-behaviour theories with elements from wider models from environmental psychology, an extensive cognitive framework that provides new and valuable insights is developed for understanding the involvement of individuals in waste prevention. The results confirm the usefulness of the theory of planned behaviour and of Schwartz's altruistic behaviour model as bases for modelling participation in waste prevention. A more elaborate integrated model of prevention was shown to be necessary for the complete analysis of attitudinal aspects associated with waste prevention. A postal survey of 158 respondents provided empirical support for eight of 12 hypotheses. The proposed structural equation indicates that personal norms and perceived behaviour control are the main predictors and that, unlike the case of recycling, subjective norms have a weak influence on WPB. It also suggests that, since social norms have not presented a direct influence, WPB is likely to be influenced by a concern for the environment and the community as well by perceptions of moral obligation and inconvenience. Results also proved that recycling and waste prevention represent different dimensions of waste management behaviour requiring particular approaches to increase individuals' engagement in future policies.

  20. Site Discharge Pollution Prevention Plan (SDPPP)

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

    SDPPP Individual Permit: Site Discharge Pollution Prevention Plan (SDPPP) The 2014 SDPPP update fully incorporates all changes made during the year and reflects changes projected for 2015. Contact Environmental Communication & Public Involvement P.O. Box 1663 MS M996 Los Alamos, NM 87545 (505) 667-0216 Email Site Discharge Pollution Prevention Plan (SDPPP) The 2014 Update to the SDPPP, Rev. 1, along with the 2013 Update to the SDPPP, Rev. 1, fulfill the requirements of Part 1.F of the

  1. Lab wins six NNSA Pollution Prevention awards

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

    Lab wins six NNSA Pollution Prevention awards Lab wins six NNSA Pollution Prevention awards The Laboratory has captured awards for projects ranging from energy savings to creating fuels from algae. March 7, 2011 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials. Los Alamos National

  2. Prevent Counter | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Prevent Counter | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA

  3. Pollution prevention opportunity assessment for organization 1700.

    SciTech Connect (OSTI)

    Gerard, Morgan Evan

    2007-06-01

    This Pollution Prevention Opportunity Assessment (PPOA) was conducted for Sandia National Laboratories/New Mexico Organization 1700 in June, 2006. The primary purpose of this PPOA is to provide recommendations to assist Organization 1700 in reducing the generation of waste and improving the efficiency of their processes and procedures. This report contains a summary of the information collected, analyses performed and recommended options for implementation. The Sandia National Laboratories Pollution Prevention staff will continue to work with Organization 1700 to implement the recommendations.

  4. Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

    SciTech Connect (OSTI)

    Pint, Bruce A.; Unocic, Kinga A.; Terrani, Kurt A.

    2015-08-01

    Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 [1-3] and this facility continues to support those efforts in the ATF community. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [4-5]. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.

  5. Type B Accident Investigation of the January 28, 2003, Fall and Injury at the Stanford Linear Accelerator Center

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by John S. Muhlestein, Director, Stanford Site Office (DOE/SC), U.S. Department of Energy.

  6. Workers at EM’s West Valley Site Surpass 1 Million Hours without Lost-Time Accident

    Broader source: Energy.gov [DOE]

    WEST VALLEY, N.Y. – EM’s cleanup contractor at the West Valley Demonstration Project (WVDP) recently marked 1 million work hours without a lost-time accident or illness.

  7. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    SciTech Connect (OSTI)

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  8. Longitudinal review of state-level accident statistics for carriers of interstate freight

    SciTech Connect (OSTI)

    Saricks, C.; Kvitek, T.

    1994-03-01

    State-level accident rates by mode of freight transport have been developed and refined for application to the US Department of Energy`s (DOE`s) environmental mitigation program, which may involve large-quantity shipments of hazardous and mixed wastes from DOE facilities. These rates reflect multi-year data for interstate-registered highway earners, American Association of Railroads member carriers, and coastal and internal waterway barge traffic. Adjustments have been made to account for the share of highway combination-truck traffic actually attributable to interstate-registered carriers and for duplicate or otherwise inaccurate entries in the public-use accident data files used. State-to-state variation in rates is discussed, as is the stability of rates over time. Computed highway rates have been verified with actual carriers of high- and low-level nuclear materials, and the most recent truck accident data have been used, to ensure that the results are of the correct order of magnitude. Study conclusions suggest that DOE use the computed rates for the three modes until (1) improved estimation techniques for highway combination-truck miles by state become available; (2) continued evolution of the railroad industry significantly increases the consolidation of interstate rail traffic onto fewer high-capacity trunk lines; or (3) a large-scale off-site waste shipment campaign is imminent.

  9. Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

    SciTech Connect (OSTI)

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln⁥Χ=-α (ΔG_rxn^° (T_C ))/(RT_C )+ÎČ were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG_rxn^° (T_C ). These models allowed an estimate of the upper bound for the reactor temperatures of T_C between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  10. SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS

    SciTech Connect (OSTI)

    Brad J. Merrill; Shannon M Bragg-Sitton

    2013-09-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) Light Water Reactor (LWR) Sustainability Program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. The Fuels Pathway within this program focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement (e.g. fully ceramic cladding). The DOE-NE Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) is also conducting research on materials for advanced, accident tolerant fuels and cladding for application in operating LWRs. To aide in this assessment, a silicon carbide (SiC) version of the MELCOR code was developed by substituting SiC in place of Zircaloy in MELCOR’s reactor core oxidation and material property routines. The purpose of this development effort is to provide a numerical capability for estimating the safety advantages of replacing Zr-alloy components in LWRs with SiC components. This modified version of the MELCOR code was applied to the Three Mile Island (TMI-2) plant accident. While the results are considered preliminary, SiC cladding showed a dramatic safety advantage over Zircaloy cladding during this accident.

  11. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect (OSTI)

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  12. Accident source terms for Light-Water Nuclear Power Plants. Final report

    SciTech Connect (OSTI)

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ``Calculation of Distance Factors for Power and Test Reactors`` which specified a release of fission products from the core to the reactor containment for a postulated accident involving ``substantial meltdown of the core``. This ``source term``, tile basis for tile NRC`s Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC`s reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ``source term`` release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ``source term`` is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it.

  13. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    SciTech Connect (OSTI)

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion.

  14. Revisiting Insights from Three Mile Island Unit 2 Postaccident Examinations and Evaluations in View of the Fukushima Daiichi Accident

    SciTech Connect (OSTI)

    Joy Rempe; Mitchell Farmer; Michael Corradini; Larry Ott; Randall Gauntt; Dana Powers

    2012-11-01

    The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized water reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.

  15. Precursors to potential severe core damage accidents, 1986: A status report: Main report and Appendixes A,B, and C

    SciTech Connect (OSTI)

    Minarick, J W; Harris, J D; Austin, P N; Cletcher, J W; Hagen, E W

    1988-05-01

    The Accident Sequence Precursor Program reviews licensee event reports of operational events that have occurred at LWRs to identify and categorize precursors to potential severe core-damage accidents. Accident sequences considered in the study are those associated with inadequate core cooling. Accident sequence precursors are events that are important elements in such sequences. Such precursors could be infrequent initiating events or equipment failures that, when coupled with one or more postulated events, could result in a plant condition with inadequate core cooling. Originally proposed in the Risk Assessment Review Group Report (Lewis Committee report) in 1978, the study - subsequently named the Accident Sequence Precursor Program - was initiated at the Nuclear Operations Analysis Center in 1979. Earlier reports by the program involved assessment of events that occurred in 1969-1981 and 1984-1985. The present report involves the assessment of events that occurred during 1986. A nuclear plant has safety systems for mitigating the consequences of accidents or off-normal initiating events that may occur during the course of plant operation. These systems are built to high-quality standards and are redundant; nonetheless, they have a nonzero probability of failing or being in a failed state when required to operate. This report uses LERs and other plant data, estimated system unavailabilities, the expected average frequency of initiating events (LOFWs, LOOPs, LOCAs), and event details to evaluate the potential impact of the following two situations.

  16. The Nuclear Accident at Three Mile Island a Practical Lesson in the Fundamental Importance of Effective Communications

    SciTech Connect (OSTI)

    DeVine Jr, J.C.

    2008-07-01

    The Three Mile Island Unit 2 (TMI-2) accident in March 1979 had a profound effect on the course of commercial nuclear generation in the United States and around the world. And while the central elements of the accident were matters of nuclear engineering, design and operations, its consequences were compounded, and in some respects superseded, by extraordinarily ineffective communications by all parties at all levels. Communications failures during the accident and its aftermath caused misunderstanding, distrust, and incorrect emergency response - and seeded or reinforced public opposition to nuclear power that persists to this day. There are communications lessons from TMI that have not yet been fully learned, and some that once were learned but are now gradually being forgotten. The more glaring TMI communications problems were in the arena of external interactions and communications among the plant owner, the Nuclear Regulatory Commission (NRC), the media, and the public. Confusing, fragmented, and contradictory public statements early in the accident, regardless of cause, undermined all possibility for reasonable discourse thereafter. And because the TMI accident was playing out on a world stage, the breakdown in public trust had long term and widespread implications. At the plant site, both TMI-2 cleanup and restart of the undamaged TMI-1 unit met with years of public and political criticism, and attendant regulatory pressure. Across the nation, public trust in nuclear power and those who operate it plummeted, unquestionably contributing to the 25+ year hiatus in new plant orders. There were other, less visible but equally important, consequences of ineffective communications at TMI. The unplanned 'precautionary' evacuation urged by the governor two days after the accident - a life changing, traumatic event for thousands of residents - was prompted primarily by misunderstandings and miscommunications regarding the condition of the plant. And today, nearly 30 years after the event, many in our nuclear industry have insufficient knowledge or regard for the underlying nuclear safety vulnerabilities revealed by the accident, in part because these have not been well explained. From this single, compelling experience, many lessons can be drawn. Some of these were recognized early and taken to heart by those who own and operate nuclear plants - but over time, respect for their importance has given way somewhat to the seemingly more urgent practicalities of plant cost, schedule and production goals. In other cases, the lessons have remained largely obscure. This paper will describe in greater detail the communications aspects of the TMI accident, lessons that can be drawn from them, and their implications on current and future nuclear facility operation. The paper reflects the author's personal, direct experience as part of the accident response team and subsequent cleanup operations at TMI. In summary: The Three Mile Accident was the most severe nuclear accident in U.S. history. It also is perhaps the most studied industrial accident of any kind in U.S. history. Exhaustive examinations of the public health consequences of the accident show convincingly that the effects of radioactivity releases, if any, were imperceptibly low. It is generally agreed, however, that there have been perceptible health consequences from the TMI-2 accident - those linked to stress. Stress to members of the public, particularly those living near the plant, was unquestionably high. And for some the combination of rumor, confusion, contradictory reports and uncertainty, all leading to an evacuation recommendation from the governor, took a toll. It could be argued that the ineffective internal and external communications during the course of the event were as influential to the outcome as the equipment and operational breakdowns that are now so well understood. And for that reason alone, this accident points out that communications capabilities - staffing, systems, facilities, training - can be as important to protection of the public, the plant an

  17. Los Alamos National Laboratory employees receive Pollution Prevention...

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

    at the Laboratory's annual Pollution Prevention Awards ceremony on Monday (April 22), Earth Day. "The Pollution Prevention Awards are the result of people taking the initiative...

  18. Prevention of Harassment (Including Sexual Harassment) and Retaliation...

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

    Prevention of Harassment (Including Sexual Harassment) and Retaliation Policy Statement Prevention of Harassment (Including Sexual Harassment) and Retaliation Policy Statement DOE...

  19. Pancreatic cancer: Pathogenesis, prevention and treatment

    SciTech Connect (OSTI)

    Sarkar, Fazlul H. Banerjee, Sanjeev; Li, Yiwei

    2007-11-01

    Pancreatic cancer is the fourth leading cause of cancer death in the United States with a very low survival rate of 5 years. To better design new preventive and/or therapeutic strategies for the fight against pancreatic cancer, the knowledge of the pathogenesis of pancreatic cancer at the molecular level is very important. It has been known that the development and the progression of pancreatic cancer are caused by the activation of oncogenes, the inactivation of tumor suppressor genes, and the deregulation of many signaling pathways among which the EGFR, Akt, and NF-{kappa}B pathways appear to be most relevant. Therefore, the strategies targeting EGFR, Akt, NF-{kappa}B, and their downstream signaling could be promising for the prevention and/or treatment of pancreatic cancer. In this brief review, we will summarize the current knowledge regarding the pathogenesis, prevention, and treatment of pancreatic cancer.

  20. Incident Prevention, Warning, and Response (IPWAR) Manual

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

    2004-09-30

    This Manual defines a structured, cohesive, and consistent process for performing incident prevention, warning, and response for DOE's Federal information systems and is consistent with the requirements of Federal laws, Executive orders, national security directives, and other regulations. The Manual also provides requirements and implementation instructions for the Department's Incident Prevention, Warning and Response process, and supplements DOE O 205.1, Department of Energy Cyber Security Management Program, dated 3-21-03. DOE N 205.17 cancels this manual. This manual cancels DOE N 205.4, Handling Cyber Security Alerts and Advisories and Reporting Cyber Security Incidents, dated 3/18/2002.

  1. Accident Investigations

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

    1995-09-29

    Cancels Para. 1 Thru 5, 6A(1) Thru (10), Both 6F (1) Thru (8) and the second misnumbered 6F, and chaps. I and II of DOE 5484.1.

  2. Accident Investigations

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

    1997-11-26

    Cancels DOE O 225.1 paragraphs 1 thru 5,6a(1) thru (10), 6b, 6d, 6f(1) thru (8), and the second misnumbered 6f; Chapters I and II of DOE O 5484.1. Cancels DOE O 225.1 in part; DOE 5484.1 in part. Canceled by DOE O 225.1B.

  3. Accident Investigations

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

    1996-07-26

    The guide explains the requirements addressed in DOe O 225.1 and provides guidance regarding acceptable methods for implementing those requirements. Canceled by DOE G 225.1A-1.

  4. Hydrogen combustion in a flat semi-confined layer with respect to the Fukushima Daiichi accident

    SciTech Connect (OSTI)

    Kuznetsov, M.; Yanez, J.; Grune, J.; Friedrich, A.; Jordan, T.

    2012-07-01

    The hydrogen accumulation at the top of containment or reactor building may occur due to an interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Daiichi). The hydrogen, released from the reactor, accumulates usually as a stratified semi-confined layer of hydrogen-air mixture. A series of large scale experiments on hydrogen combustion and explosion in a semi-confined layer of uniform and non-uniform hydrogen-air mixtures in presence of obstructions or without them was performed at the Karlsruhe Inst. of Technology (KIT). Different flame propagation regimes from slow subsonic to relative fast sonic flames and then to the detonations were experimentally investigated in different geometries and then simulated with COMSD code with respect to evaluate amount of burnt hydrogen taken place during the Fukushima Daiichi Accident (FDA). The experiments were performed in a horizontal semi-confined layer with dimensions of 9x3x0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 0-34 vol. % without or with a gradient of 0-60 vol. %H{sub 2}/m. Effects of hydrogen concentration gradient, thickness of the layer, geometry of the obstructions, average and maximum hydrogen concentration on flame propagation regimes were investigated with respect to evaluate the maximum pressure loads of internal structures. Blast wave strength and dynamics of propagation after explosion of the layer of hydrogen-air mixture was numerically simulated to reproduce the hydrogen explosion process during the Fukushima Daiichi Accident. (authors)

  5. Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

    SciTech Connect (OSTI)

    Zhao, Haihua; Zhang, Hongbin; Zou, Ling; Anders, David; Martineau, Richard

    2014-10-01

    The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC) system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.

  6. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    SciTech Connect (OSTI)

    Wang, T.-C.; Wang, S.-J.; Teng, J.-T

    2005-05-15

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP.

  7. Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

    SciTech Connect (OSTI)

    Miller, W.J.

    1980-05-01

    This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

  8. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect (OSTI)

    Rebak, Raul B.

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 ”m. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

  9. Public meetings on radiation and its health effects caused by the Fukushima nuclear accident

    SciTech Connect (OSTI)

    Sugiyama, K.; Ayame, J.; Takashita, H.; Yamamoto, R.

    2013-07-01

    The Japan Atomic Energy Agency (JAEA) has held public meetings on radiation and its health effects mainly for parents of students in kindergartens, elementary schools, and junior high schools in Fukushima and Ibaraki prefectures after the Fukushima nuclear accident. These meetings are held based on our experience of practicing risk communication activities for a decade in JAEA with local residents. By analyzing questionnaires collected after the meetings, we confirmed that interactive communication is effective in increasing participants' understanding and in decreasing their anxiety. Most of the participants answered that they understood the contents and that it eased their mind. (authors)

  10. Nuclear reactor accidents: Chernobyl, TMI, and Windscale. (Latest citations from Pollution abstracts). Published Search

    SciTech Connect (OSTI)

    1995-12-01

    The bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout; the resultant runoff into rivers, lakes, and the sea; the radiation effects on people; and the transfrontier radioactive contamination of the environment. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  11. Fission product source terms for the LWR loss-of-coolant accident

    SciTech Connect (OSTI)

    Lorenz, R.A.; Collins, J.L.; Malinauskas, A.P.

    1980-07-01

    Models for cesium and iodine release from light-water reactor (LWR) fuel rods failed in steam were formulated based on experimental fission product release data from several types of failed LWR fuel rods. The models were applied to a pressurized water reactor (PWR) undergoing a hypothetical loss-of-coolant accident (LOCA) temperature transient. Calculated total iodine and cesium releases from the fuel rods were 0.053 and 0.025% of the total reactor inventories of these elements, respectively, with most of the release occurring at the time of rupture. These values are approximately two orders of magnitude less than releases used in WASH-1400, the Reactor Safety Study.

  12. SACO-1: a fast-running LMFBR accident-analysis code

    SciTech Connect (OSTI)

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  13. Type A Accident Investigation of the July 15, 2004, Hanford 200 East Area

    Energy Savers [EERE]

    Fall Fatality | Department of Energy July 15, 2004, Hanford 200 East Area Fall Fatality Type A Accident Investigation of the July 15, 2004, Hanford 200 East Area Fall Fatality August 1, 2004 On the morning of Thursday, July 15, 2004, a non-government contractor employee of All Mobile Transporting & Repairs (AMTR) was found motionless at the bottom of a ladder with a serious head injury. It was believed the employee had been standing on the ladder, using a battery-powered drill, to remove

  14. Type B Accident Investigation Board Report on the Head Injury to a Miner at

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

    the Waste Isolation Pilot Plant, Carlsbad, New Mexico - August 25, 2004 | Department of Energy on the Head Injury to a Miner at the Waste Isolation Pilot Plant, Carlsbad, New Mexico - August 25, 2004 Type B Accident Investigation Board Report on the Head Injury to a Miner at the Waste Isolation Pilot Plant, Carlsbad, New Mexico - August 25, 2004 October 15, 2004 On August 25, 2004, an employee of Washington TRU Solution, LLC (WTS) sustained a head injury when he was struck by a C-clamp and

  15. Type B Accident Investigation of the Acid Vapor Inhalation on June 7, 2005,

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

    in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory | Department of Energy Acid Vapor Inhalation on June 7, 2005, in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory Type B Accident Investigation of the Acid Vapor Inhalation on June 7, 2005, in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory Setpember 27, 2005 During the afternoon of June 7, 2005, two postdoctoral (postdoc) employees working at the Los Alamos National Laboratory

  16. Type B Accident Investigation of the January 10, 2006, Flash Fire and

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

    Injury at the Savannah River National Laboratory | Department of Energy January 10, 2006, Flash Fire and Injury at the Savannah River National Laboratory Type B Accident Investigation of the January 10, 2006, Flash Fire and Injury at the Savannah River National Laboratory February 1, 2006 On January 10, 2006, at approximately 7:47 a.m., a first-line manager (FLM) at the Savannah River National Laboratory (SRNL) received first- and second-degree burns to his head, face, neck, and left hand

  17. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple Uptake

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

    Event at the Pu Facility, Los Alamos National Laboratory | Department of Energy on the August 5, 2003, Pu-238 Multiple Uptake Event at the Pu Facility, Los Alamos National Laboratory Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple Uptake Event at the Pu Facility, Los Alamos National Laboratory December 1, 2003 On August 5, 2003, a release of plutonium-238 occurred in a storage room at the Plutonium Facility, Los Alamos National Laboratory, resulting in radiation doses to

  18. Guide to Preventing Computer Software Piracy

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

    2001-07-12

    Guide to Preventing Computer Software Piracy It is the intent of the Department of Energy (DOE) to issue guidance in accordance with Federal CIO Council recommendations and in compliance with Executive Order 13103. The guidance in this document is based on the CIO Council's recommendations in reference to computer software piracy, and applies to all DOE elements. Canceled by DOE N 205.18

  19. Pollution Prevention, Waste Reduction, and Recycling

    Broader source: Energy.gov [DOE]

    The DOE Pollution Prevention, Waste Reduction and Recycling Program supports Executive Order (E.O.) 13693, Planning for Federal Sustainability in the Next Decade; DOE Order 436.1, Department Sustainability; and, the Department’s Strategic Sustainability Performance Plan (SSPP). Under this Program, Headquarters assists DOE Field and Program Offices in the implementation of the specific goals identified in both of the Orders.

  20. Policy Flash 2013-28 Implementation of Indian Energy Preference Provision in EP Act 2005 and Acquisition Letter 2013-02

    Broader source: Energy.gov [DOE]

    Attached is Policy Flash 2013-28 Implementation of Indian Energy Preference Provision in EP Act 2005 and Acquisition Letter 2013-02

  1. Type B Accident Investigation Board Report on the May 7, 1997, Worker Injury at the Hanford Site Canister Storage Building Construction Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Michael S. Cowan, Chief Program Officer, Western Area Power Administration.

  2. Type B Accident Investigation of the Mineral Oil Leak Discovered on January 8, 2001, Resulting in Property Damage at the Atlas Facility, Los Alamos National Laboratory

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Acting Chief Operating Officer for Defense Programs, Ralph E. Erickson.

  3. Type B Accident Investigation Board Report on the September 15, 1997, Drum Explosion at Building C-746-Q, Paducah Gaseous Diffusion Plant, Paducah, Kentucky

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board (Board) appointed by James C. Hall, Manager, Oak Ridge Operations.

  4. Safety Analysis: Evaluation of Accident Risks in the Transporation of Hazardous Materials by Truck and Rail at the Savannah River Plant

    SciTech Connect (OSTI)

    Blanchard, A.

    1999-04-15

    This report presents an analysis of the consequences and risks of accidents resulting from hazardous material transportation at the Savannah River Plant.

  5. Type B Accident Investigation Board Report on the November 1, 1999, Construction Injury at the Monticello Mill Tailings Remedial Action Site, Monticello, Utah

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B accident investigation board appointed by R. E. Glass, Manager, Albuquerque Operations Office.

  6. Type B Accident Investigation Board Report on the October 22, 1997, Electrical Arc Blast at Building F-Zero Fermi National Accelerator Laboratory, Batavia, Illinois

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Cherri J. Langenfeld, Manager, Chicago Operations Office, U.S. Department of Energy.

  7. Type B Accident Investigation Of The February 25, 2009 Injury To A Passenger In An Electric Cart At The Waste Isolation Pilot Plant, Carlsbad, New Mexico

    Broader source: Energy.gov [DOE]

    This report is a product of an Accident Investigation Board appointed by David C. Moody, Manager, Carlsbad Field Office, Department of Energy, on March 4, 2009.

  8. Type B Accident Investigation Board Report on the September 1, 1999, Plutonium Intakes at the Savannah River Site FB-Line

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Greg Rudy, Manager, Savannah River Operations Office, U.S. Department of Energy.

  9. Type B Accident Investigation Board Report on the November 17, 1997, Chiller Line Rupture at Technical Area 35, Building 27, Los Alamos National Laboratory

    Broader source: Energy.gov [DOE]

    This report is a product of an accident investigation board appointed by Bruce G. Twining, Manager, Albuquerque Operations Office, Department of Energy.

  10. Type A Accident Investigation Board Report on the July 11, 1996, Electrical Shock at Technical Area 53, Building MPF-14, Los Alamos National Laboratory

    Broader source: Energy.gov [DOE]

    This report is an independent product of an electrical shock accident investigation board appointed by Bruce G. Twining, Manager, Albuquerque Operations Office, Department of Energy.

  11. 2016 National Fall Prevention Campaign | Department of Energy

    Energy Savers [EERE]

    National Fall Prevention Campaign 2016 National Fall Prevention Campaign March 17, 2016 - 9:07am Addthis 2016 National Fall Prevention Campaign As part of a Fall Prevention event initiated by the Occupational Safety and Health Administration (OSHA), the 3rd Annual National Fall Prevention Campaign will take place on May 2-6. This event is a nationwide voluntary effort to remind and educate employers and workers in the construction industry of the serious dangers regarding falls from elevated

  12. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    SciTech Connect (OSTI)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  13. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    SciTech Connect (OSTI)

    J.R. Smotrel; H. Marr; M.J. Anderson

    2001-04-05

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities.

  14. Risk communication with Fukushima residents affected by the Fukushima Daiichi accident at whole-body counting

    SciTech Connect (OSTI)

    Gunji, I.; Furuno, A.; Yonezawa, R.; Sugiyama, K.

    2013-07-01

    After the Tokyo Electric Power Company (TEPCO) Fukushima Daiichi nuclear power plant accident, the Tokai Research and Development Center of the Japan Atomic Energy Agency (JAEA) have had direct dialogue as risk communication with Fukushima residents who underwent whole-body counting examination (WBC). The purpose of the risk communication was to exchange information and opinions about radiation in order to mitigate Fukushima residents' anxiety and stress. Two kinds of opinion surveys were performed: one survey evaluated residents' views of the nuclear accident itself and the second survey evaluated the management of WBC examination as well as the quality of JAEA's communication skills on risks. It appears that most Fukushima residents seem to have reduced their anxiety level after the direct dialogue. The results of the surveys show that Fukushima residents have the deepest anxiety and concern about their long-term health issues and that they harbor anger toward the government and TEPCO. On the other hand, many WBC patients and patients' relatives have expressed gratitude for help in reducing their feelings of anxiety.

  15. LLNL Results from CALIBAN-PROSPERO Nuclear Accident Dosimetry Experiments in September 2014

    SciTech Connect (OSTI)

    Lobaugh, M. L.; Hickman, D. P.; Wong, C. W.; Wysong, A. R.; Merritt, M. J.; Heinrichs, D. P.; Topper, J. D.

    2015-05-21

    Lawrence Livermore National Laboratory (LLNL) uses thin neutron activation foils, sulfur, and threshold energy shielding to determine neutron component doses and the total dose from neutrons in the event of a nuclear criticality accident. The dosimeter also uses a DOELAP accredited Panasonic UD-810 (Panasonic Industrial Devices Sales Company of America, 2 Riverfront Plaza, Newark, NJ 07102, U.S.A.) thermoluminescent dosimetery system (TLD) for determining the gamma component of the total dose. LLNL has participated in three international intercomparisons of nuclear accident dosimeters. In October 2009, LLNL participated in an exercise at the French Commissariat à l’énergie atomique et aux énergies alternatives (Alternative Energies and Atomic Energy Commission- CEA) Research Center at Valduc utilizing the SILENE reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison at CEA Valduc, this time with exposures at the CALIBAN reactor (Hickman et al. 2011). This paper discusses LLNL’s results of a third intercomparison hosted by the French Institut de Radioprotection et de Sûreté Nucléaire (Institute for Radiation Protection and Nuclear Safety- IRSN) with exposures at two CEA Valduc reactors (CALIBAN and PROSPERO) in September 2014. Comparison results between the three participating facilities is presented elsewhere (Chevallier 2015; Duluc 2015).

  16. DYNAMIC ANALYSIS OF HANFORD UNIRRADIATED FUEL PACKAGE SUBJECTED TO SEQUENTIAL LATERAL LOADS IN HYPOTHETICAL ACCIDENT CONDITIONS

    SciTech Connect (OSTI)

    Wu, T

    2008-04-30

    Large fuel casks present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. This paper presents a numerical technique for evaluating the dynamic responses of large fuel casks subjected to sequential HAC loading. A nonlinear dynamic analysis was performed for a Hanford Unirradiated Fuel Package (HUFP) [1] to evaluate the cumulative damage after the hypothetical accident Conditions of a 30-foot lateral drop followed by a 40-inch lateral puncture as specified in 10CFR71. The structural integrity of the containment vessel is justified based on the analytical results in comparison with the stress criteria, specified in the ASME Code, Section III, Appendix F [2], for Level D service loads. The analyzed cumulative damages caused by the sequential loading of a 30-foot lateral drop and a 40-inch lateral puncture are compared with the package test data. The analytical results are in good agreement with the test results.

  17. Interactive Rapid Dose Assessment Model (IRDAM): reactor-accident assessment methods. Vol. 2

    SciTech Connect (OSTI)

    Poeton, R.W.; Moeller, M.P.; Laughlin, G.J.; Desrosiers, A.E.

    1983-05-01

    As part of the continuing emphasis on emergency preparedness, the US Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). This system, the Interactive Rapid Dose Assessment Model (IRDAM) is a micro-computer based program for rapidly assessing the radiological impact of accidents at nuclear power plants. This document describes the technical bases for IRDAM including methods, models and assumptions used in calculations. IRDAM calculates whole body (5-cm depth) and infant thyroid doses at six fixed downwind distances between 500 and 20,000 meters. Radionuclides considered primarily consist of noble gases and radioiodines. In order to provide a rapid assessment capability consistent with the capacity of the Osborne-1 computer, certain simplifying approximations and assumptions are made. These are described, along with default values (assumptions used in the absence of specific input) in the text of this document. Two companion volumes to this one provide additional information on IRDAM. The user's Guide (NUREG/CR-3012, Volume 1) describes the setup and operation of equipment necessary to run IRDAM. Scenarios for Comparing Dose Assessment Models (NUREG/CR-3012, Volume 3) provides the results of calculations made by IRDAM and other models for specific accident scenarios.

  18. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

    2015-06-21

    In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U3Si2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U3Si2 has been optimized and high phase purity U3Si2 has been successfully produced. Results are presented from sintering studies and microstructural examinationsmore » that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

  19. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

    SciTech Connect (OSTI)

    Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

    2015-06-21

    In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U3Si2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U3Si2 has been optimized and high phase purity U3Si2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.

  20. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

    SciTech Connect (OSTI)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harrison, J.D.; Harper, F.T.; Hora, S.C.

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.