National Library of Energy BETA

Sample records for final safety analysis

  1. FFTF Final Safety Analysis Report Amendment 82

    SciTech Connect (OSTI)

    DAUTEL, W.A.

    2003-03-01

    This is the Final Safety Analysis Report (FSAR) Amendment 82 for incorporation into the Fast Flux Test Facility (FFTF) FSAR set assigned to you. This page change amendment incorporates changes previously approved by the U.S. Department of Energy, Richland Operations Office. This amendment provides updates to the FSAR to facilitate FFTF shutdown and deactivation. Among the changes are the following: Chapter 11 is updated to describe upgrades to the Solid Waste Cask. Additional fuel handling accidents are added to Chapter 15. Appendix G is revised to clearly identify systems and their safety functions. Appendix H is revised to remove the discussion of material that has been removed from the Interim Storage Area. Appendix I is revised to provide a general description of liquid metal removal from FFTF. Other changes include minor technical updates from the FSAR annual review and editorial and procedure references.

  2. FFTF Final Safety Analysis Report Amendment 81 [SEC 1 & 2

    SciTech Connect (OSTI)

    DAUTEL, W.A.

    2002-01-10

    Since the last reactor operation of FFTF in March of 1992, the FFTF has either been in a programmatic status of Standby or Shutdown. The facility hazards have decreased markedly. Rather than making extensive Final Safety Analysis Report (FSAR) changes, Appendix G was prepared to reflect the design and operation during Standby or Shutdown. Appendix G describes the application of the entire FSAR for the current configuration, accounting for the natural reduction in hazards and new system configurations associated with Standby/Shutdown. The technical system chapters and the safety analysis chapter of the FSAR describe how the design and operation fulfilled the requirements necessary to support reactor operation; this information is retained for design basis and historical information. This Final Safety Analysis Report (FSAR) is submitted per the requirements of Paragraph 014, Energy Research and Development Administration (ERDA) Manual Chapter 0540, ''Safety of ERDA-Owned Reactors.'' This FSAR and its supporting documentation provide a complete description and safety evaluation of the site, plant design, normal and emergency operations, potential accidents and predicted consequences of such accidents, and the means that will prevent such accidents and/or reduce their consequences to an acceptable level.

  3. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    SciTech Connect (OSTI)

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  4. TA-55 Final Safety Analysis Report Comparison Document and DOE Safety Evaluation Report Requirements

    SciTech Connect (OSTI)

    Alan Bond

    2001-04-01

    This document provides an overview of changes to the currently approved TA-55 Final Safety Analysis Report (FSAR) that are included in the upgraded FSAR. The DOE Safety Evaluation Report (SER) requirements that are incorporated into the upgraded FSAR are briefly discussed to provide the starting point in the FSAR with respect to the SER requirements.

  5. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    SciTech Connect (OSTI)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-11-09

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG&G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort.

  6. 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.

  7. Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

    SciTech Connect (OSTI)

    ULLAH, M K

    2001-02-26

    The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

  8. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

    SciTech Connect (OSTI)

    CARRELL, R D

    2002-07-16

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

  9. Final safety analysis report for the Ground Test Accelerator (GTA), Phase 2

    SciTech Connect (OSTI)

    1994-10-01

    This document is the second volume of a 3 volume safety analysis report on the Ground Test Accelerator (GTA). The GTA program at the Los Alamos National Laboratory (LANL) is the major element of the national Neutral Particle Beam (NPB) program, which is supported by the Strategic Defense Initiative Office (SDIO). A principal goal of the national NPB program is to assess the feasibility of using hydrogen and deuterium neutral particle beams outside the Earth`s atmosphere. The main effort of the NPB program at Los Alamos concentrates on developing the GTA. The GTA is classified as a low-hazard facility, except for the cryogenic-cooling system, which is classified as a moderate-hazard facility. This volume consists of failure modes and effects analysis; accident analysis; operational safety requirements; quality assurance program; ES&H management program; environmental, safety, and health systems critical to safety; summary of waste-management program; environmental monitoring program; facility expansion, decontamination, and decommissioning; summary of emergency response plan; summary plan for employee training; summary plan for operating procedures; glossary; and appendices A and B.

  10. Final safety analysis report for the Galileo Mission: Volume 1, Reference design document

    SciTech Connect (OSTI)

    Not Available

    1988-05-01

    The Galileo mission uses nuclear power sources called Radioisotope Thermoelectric Generators (RTGs) to provide the spacecraft's primary electrical power. Because these generators contain nuclear material, a Safety Analysis Report (SAR) is required. A preliminary SAR and an updated SAR were previously issued that provided an evolving status report on the safety analysis. As a result of the Challenger accident, the launch dates for both Galileo and Ulysses missions were later rescheduled for November 1989 and October 1990, respectively. The decision was made by agreement between the DOE and the NASA to have a revised safety evaluation and report (FSAR) prepared on the basis of these revised vehicle accidents and environments. The results of this latest revised safety evaluation are presented in this document (Galileo FSAR). Volume I, this document, provides the background design information required to understand the analyses presented in Volumes II and III. It contains descriptions of the RTGs, the Galileo spacecraft, the Space Shuttle, the Inertial Upper Stage (IUS), the trajectory and flight characteristics including flight contingency modes, and the launch site. There are two appendices in Volume I which provide detailed material properties for the RTG.

  11. Final safety analysis report for the Ground Test Accelerator (GTA), Phase 2

    SciTech Connect (OSTI)

    1994-10-01

    This document is the third volume of a 3 volume safety analysis report on the Ground Test Accelerator (GTA). The GTA program at the Los Alamos National Laboratory (LANL) is the major element of the national Neutral Particle Beam (NPB) program, which is supported by the Strategic Defense Initiative Office (SDIO). A principal goal of the national NPB program is to assess the feasibility of using hydrogen and deuterium neutral particle beams outside the Earth`s atmosphere. The main effort of the NPB program at Los Alamos concentrates on developing the GTA. The GTA is classified as a low-hazard facility, except for the cryogenic-cooling system, which is classified as a moderate-hazard facility. This volume consists of appendices C through U of the report

  12. Final safety analysis report for the Ground Test Accelerator (GTA), Phase 2

    SciTech Connect (OSTI)

    1994-10-01

    This document is the first volume of a 3 volume safety analysis report on the Ground Test Accelerator (GTA). The GTA program at the Los Alamos National Laboratory (LANL) is the major element of the national Neutral Particle Beam (NPB) program, which is supported by the Strategic Defense Initiative Office (SDIO). A principal goal of the national NPB program is to assess the feasibility of using hydrogen and deuterium neutral particle beams outside the Earth`s atmosphere. The main effort of the NPB program at Los Alamos concentrates on developing the GTA. The GTA is classified as a low-hazard facility, except for the cryogenic-cooling system, which is classified as a moderate-hazard facility. This volume consists of an introduction, summary/conclusion, site description and assessment, description of facility, and description of operation.

  13. Full-length high-temperature severe fuel damage test No. 2. Final safety analysis

    SciTech Connect (OSTI)

    Hesson, G.M.; Lombardo, N.J.; Pilger, J.P.; Rausch, W.N.; King, L.L.; Hurley, D.E.; Parchen, L.J.; Panisko, F.E.

    1993-09-01

    Hazardous conditions associated with performing the Full-Length High- Temperature (FLHT). Severe Fuel Damage Test No. 2 experiment have been analyzed. Major hazards that could cause harm or damage are (1) radioactive fission products, (2) radiation fields, (3) reactivity changes, (4) hydrogen generation, (5) materials at high temperature, (6) steam explosion, and (7) steam pressure pulse. As a result of this analysis, it is concluded that with proper precautions the FLHT- 2 test can be safely conducted.

  14. Waste Receiving and Processing (WRAP) Facility Final Safety Analysis Report (FSAR)

    SciTech Connect (OSTI)

    TOMASZEWSKI, T.A.

    2000-04-25

    The Waste Receiving and Processing Facility (WRAP), 2336W Building, on the Hanford Site is designed to receive, confirm, repackage, certify, treat, store, and ship contact-handled transuranic and low-level radioactive waste from past and present U.S. Department of Energy activities. The WRAP facility is comprised of three buildings: 2336W, the main processing facility (also referred to generically as WRAP); 2740W, an administrative support building; and 2620W, a maintenance support building. The support buildings are subject to the normal hazards associated with industrial buildings (no radiological materials are handled) and are not part of this analysis except as they are impacted by operations in the processing building, 2336W. WRAP is designed to provide safer, more efficient methods of handling the waste than currently exist on the Hanford Site and contributes to the achievement of as low as reasonably achievable goals for Hanford Site waste management.

  15. Final safety analysis report for the Galileo mission: Volume 3 (Book 2), Nuclear risk analysis document: Appendices: Revision 1

    SciTech Connect (OSTI)

    Not Available

    1989-01-25

    It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequence analyses can be used in several ways. One way is to identify the potential range of consequences which might have to be dealt with if there were to be an accident with a release of fuel, so as to assure that, given such an accident, the health and safety of the public will be reasonably protected. Another use of the information, in conjunction with accident and release probabilities, is to estimate the risks associated with the mission. That is, most space launches occur without incident. Given an accident, the most probable result relative to the RTGs is complete containment of the radioactive material. Only a small fraction of accidents might result in a release of fuel and subsequent radiological consequences. The combination of probability with consequence is risk, which can be compared to other human and societal risks to assure that no undue risks are implied by undertaking the mission. Book 2 contains eight appendices.

  16. Pantex Plant final safety analysis report, Zone 4 magazines. Staging or interim storage for nuclear weapons and components: Issue D

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    This Safety Analysis Report (SAR) contains a detailed description and evaluation of the significant environmental, safety, and health (ES&H) issues associated with the operations of the Pantex Plant modified-Richmond and steel arch construction (SAC) magazines in Zone 4. It provides (1) an overall description of the magazines, the Pantex Plant, and its surroundings; (2) a systematic evaluations of the hazards that could occur as a result of the operations performed in these magazines; (3) descriptions and analyses of the adequacy of the measures taken to eliminate, control, or mitigate the identified hazards; and (4) analyses of potential accidents and their associated risks.

  17. K Basin safety analysis

    SciTech Connect (OSTI)

    Porten, D.R.; Crowe, R.D.

    1994-12-16

    The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall.

  18. CRAD, Nuclear Safety Delegations for Documented Safety Analysis...

    Office of Environmental Management (EM)

    Nuclear Safety Delegations for Documented Safety Analysis Approval - January 8, 2015 (EA CRAD 31-09, Rev. 0) CRAD, Nuclear Safety Delegations for Documented Safety Analysis ...

  19. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    SciTech Connect (OSTI)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  20. Technical and Analytical Support Services to the Office of Environmental Analysis, Office of Environment, Safety and Health. Final report

    SciTech Connect (OSTI)

    1995-02-01

    The primary purpose of this contract was to provide technical analyses, studies, and reviews related to land use/water issues and energy resource development in support of the activities of the Office of Environmental Analysis, Office of Environment, Safety and Health. Tasks under this contract included: Issue Papers. Energetics provided issue papers on a number of specific energy and environmental issue areas. Each issue paper consisted of a systematic review and analysis of major factors (technical, legal, environmental, economic, energy, health and social) that could enter into DOE`s environmental/energy policy decisions; Special Analyses. Energetics conducted special in-depth technical analyses as requested by the Contracting Officer`s Technical Representative (COTR); and Critical Review and Evaluation of Program Reports. Energetics performed critical reviews of a number of technical reports arising from DOE program activities. These documents included issue papers and reports resulting from special technical analyses of specific issues, technologies, or broad areas of concern. Reviews focused on both the technical and programmatic impact of the report. Energetics made recommendations and gave input to assist DOE in determining the environmental impacts of energy policies and projects.

  1. Analysis of Integrated Safety Management at the Activity Level: Work

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

    Planning and Control, Final Report | Department of Energy Analysis of Integrated Safety Management at the Activity Level: Work Planning and Control, Final Report Analysis of Integrated Safety Management at the Activity Level: Work Planning and Control, Final Report May 15, 2013 Presenter: Stephen L. Domotor, Director, Office of Analysis, Office of Health, Safety and Security Topic: On August 28, 2012, the Defense Nuclear Facilities Safety Board (DNFSB or "Board") wrote to the

  2. Analysis of Integrated Safety Management at the Activity Level...

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

    Analysis of Integrated Safety Management at the Activity Level: Work Planning and Control, Final Report May 15, 2013 Presenter: Stephen L. Domotor, Director, Office of Analysis, ...

  3. CRAD, Facility Safety- Documented Safety Analysis

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Documented Safety Analysis.

  4. Documented Safety Analysis (DSA) Approved

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

    The latest DSA was developed based on the new DOE standard DOE-STD-3009-2014, Preparation of Non-Reactor Nuclear Facility Documented Safety Analysis, and defines the safety ...

  5. FAQS Gap Analysis Qualification Card - Occupational Safety |...

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

    Occupational Safety FAQS Gap Analysis Qualification Card - Occupational Safety Functional Area Qualification Standard Gap Analysis Qualification Cards outline the differences ...

  6. safety analysis report

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

    Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering ...

  7. Final Hazard Categorization and Auditable Safety Analysis for the Remediation of the 118-D-1, 118-D-2, 118-D-3, 118-H-1, 118-H-2 and 118-H-3 Solid Waste Burial Grounds

    SciTech Connect (OSTI)

    T. J. Rodovsky

    2006-03-01

    This report presents the initial hazard categorization, final hazard categorization and auditable safety analysis for the remediation of the 118-D-1, 118-D-2, and 118-D-3 Burial Grounds located within the 100-D/DR Area of the Hanford Site and the 118-H-1, 118-H-2, and 118-H-3 Burial Grounds located within the 100-H Area of the Hanford Site.

  8. Safety analysis report: packages. GPHS shipping package supplement 2 to the PISA shipping package (packaging of fissile and other radioactive materials). Final report

    SciTech Connect (OSTI)

    Chalfant, G. G.

    1981-06-01

    Safety Analysis Report DPST-78-124-1 is amended to permit shipment of 6 General Purpose Heat Source (GPHS) capsules (max.). Each capsule contains an average of 2330 curies of /sup 238/Pu, and each pair of capsules is contained in a welded stainless steel primary containment vessel, all of which are doubly contained in a flanged secondary containment vessel. This is in addition to the forms discussed in DPST-78-124-1 and Supplement 1.

  9. Corporate Analysis of DOE Safety Performance

    Broader source: Energy.gov [DOE]

    The Office of Environment, Health, Safety and Security (EHSS), Office of Analysis develops analysis tools and performance dashboards, and conducts analysis of DOE safety performance corporately and on a variety of specific environment, safety and health topics.

  10. Events Beyond Design Safety Basis Analysis

    Broader source: Energy.gov [DOE]

    This Safety Alert provides information on a safety concern related to the identification and mitigation of events that may fall outside those analyzed in the documented safety analysis. [Safety Bulletin 2011-01

  11. Improving the safety of LWR power plants. Final report

    SciTech Connect (OSTI)

    Not Available

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs (improving or maintaining level of safety with simpler systems or in a more cost-effective manner).

  12. Waste Isolation Pilot Plant Safety Analysis Report

    SciTech Connect (OSTI)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  13. CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval

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

    - January 8, 2015 (EA CRAD 31-09, Rev. 0) | Department of Energy Nuclear Safety Delegations for Documented Safety Analysis Approval - January 8, 2015 (EA CRAD 31-09, Rev. 0) CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval - January 8, 2015 (EA CRAD 31-09, Rev. 0) January 8, 2015 Nuclear Safety Delegations for Documented Safety Analysis Approval (EA CRAD 31-09, Rev. 0) This Criteria Review and Approach Document (EA CRAD 31-09, Rev. 0) provides objectives, criteria,

  14. Preparation of Documented Safety Analysis for Decommissioning...

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

    ... for the purpose of establishing functional and performance requirements for safety ... and long-term stewardship. DOE O 430.1B, Attachment 3 Documented Safety Analysis (DSA). ...

  15. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    SciTech Connect (OSTI)

    E.N. Lindner

    2004-12-03

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  16. Assessment of technologies for hazardous waste site remediation: Non-treatment technologies and pilot scale facility implementation -- excavation -- storage technology -- safety analysis and review statement. Final report

    SciTech Connect (OSTI)

    Johnson, H.R.; Overbey, W.K. Jr.; Koperna, G.J. Jr.

    1994-02-01

    The purpose of this study is to assess the state-of-the-art of excavation technology as related to environmental remediation applications. A further purpose is to determine which of the excavation technologies reviewed could be used by the US Corp of Engineers in remediating contaminated soil to be excavated in the near future for construction of a new Lock and Dam at Winfield, WV. The study is designed to identify excavation methodologies and equipment which can be used at any environmental remediation site but more specifically at the Winfield site on the Kanawha River in Putnam County, West Virginia. A technical approach was determined whereby a functional analysis was prepared to determine the functions to be conducted during the excavation phase of the remediation operations. A number of excavation technologies were identified from the literature. A set of screening criteria was developed that would examine the utility and ranking of the technologies with respect to the operations that needed to be conducted at the Winfield site. These criteria were performance, reliability, implementability, environmental safety, public health, and legal and regulatory compliance. The Loose Bulk excavation technology was ranked as the best technology applicable to the Winfield site. The literature was also examined to determine the success of various methods of controlling fugitive dust. Depending upon any changes in the results of chemical analyses, or prior remediation of the VOCs from the vadose zone, consideration should be given to testing a new ``Pneumatic Excavator`` which removes the VOCs liberated during the excavation process as they outgas from the soil. This equipment however would not be needed on locations with low levels of VOC emissions.

  17. Safety System Oversight Staffing Analysis - Example | Department...

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

    Staffing Analysis - Example Safety System Oversight Staffing Analysis - Example SSO Alternate Staffing Analysis - Example Filled Spreadsheet. (23 KB) More Documents & Publications ...

  18. Safety study application guide. Safety Analysis Report Update Program

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly {open_quotes}low{close_quotes}) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, {open_quotes}Technical Safety Requirements,{close_quotes} and 5480.23, {open_quotes}Nuclear Safety Analysis Reports.{close_quotes} A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis.

  19. Monthly Analysis of Electrical Safety Occurrences - February...

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

    information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - February 2011 More Documents & Publications Monthly Analysis of...

  20. Safety System Oversight Staffing Analysis (Instructions, Blank...

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

    Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and Example Sheet) This Staffing Analysis calculation is completed using an Excel worksheet. Information ...

  1. System Safety Common Cause Analysis

    Energy Science and Technology Software Center (OSTI)

    1992-03-10

    The COMCAN fault tree analysis codes are designed to analyze complex systems such as nuclear plants for common causes of failure. A common cause event, or common mode failure, is a secondary cause that could contribute to the failure of more than one component and violates the assumption of independence. Analysis of such events is an integral part of system reliability and safety analysis. A significant common cause event is a secondary cause common tomore » all basic events in one or more minimal cut sets. Minimal cut sets containing events from components sharing a common location or a common link are called common cause candidates. Components share a common location if no barrier insulates any one of them from the secondary cause. A common link is a dependency among components which cannot be removed by a physical barrier (e.g.,a common energy source or common maintenance instructions).« less

  2. Assuring the quality of safety analyses and safety analysis documentation

    SciTech Connect (OSTI)

    J. E. Johnson

    2000-05-03

    Planning, preparation, and submittal of safety analysis reports might be pursued in a manner similar to a quality-related procurement, where customer needs, expectations and acceptance criteria are established in advance. Then the product/service provider, the contractor, should apply various quality control processes to assure the desired characteristics of the product safety analysis documents. Improving the quality and acceptability to DOE of safety documents at first submittal should result in a more expeditious DOE review and approval process, thereby reducing costs of network and recycle through reviews.

  3. Assuring the Quality of Safety Analyses and Safety Analysis Documentation

    SciTech Connect (OSTI)

    Johnson, John Edwin

    2000-05-01

    Planning, preparation, and submittal of safety analysis reports might be pursued in a manner similar to a quality-related procurement, where customer needs, expectations and acceptance criteria are established in advance. Then the product/service provider, the contractor, should apply various quality control processes to assure the desired characteristics of the product safety analysis documents. Improving the quality and acceptability to DOE of safety documents at first submittal should result in a more expeditious DOE review and approval process, thereby reducing costs of network and recycle through reviews.

  4. Solid waste burial grounds interim safety analysis

    SciTech Connect (OSTI)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  5. Radiological Safety Analysis Code System.

    Energy Science and Technology Software Center (OSTI)

    2009-12-22

    Version 03 RSAC-6.2 can be used to model complex accidents and radiological consequences to individuals from the release of radionuclides to the atmosphere. A user can generate a fission product inventory; decay and ingrow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Doses are calculated through the inhalation, immersion, ground surface and ingestion pathways. New to RSAC-6.2 are the abilitiesmore » to calculate inhalation from release to a room, inhalation from resuspension of activities, and a new model for dry deposition. Doses can now be calculated as close as 10 meters from the release point. RSAC-6.2 has been subjected to extensive independent verification and validation for use in performing safety-related dose calculations to support safety analysis reports. WinRP 2.0, a windows based overlay to RSAC-6.2, assists users in creating and running RSAC-6.2 input files. RSAC-6, Rev. 6.2 (03/11/02) corrects an earlier issue with RSAC-6, compiled with F77L-EM/32 Fortran 77 Version 5.10, which would not allow the executable to run with XP or VISTA Windows operating systems. Because this version is still in use at some facilities, it is being released through RSICC in addition to the new RSAC 7 (CCC-761).« less

  6. Final Report - Hydrogen Delivery Infrastructure Options Analysis...

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

    - Hydrogen Delivery Infrastructure Options Analysis Final Report - Hydrogen Delivery Infrastructure Options Analysis This report, by the Nexant team, documents an in-depth analysis ...

  7. Hot Cell Facility (HCF) Safety Analysis Report

    SciTech Connect (OSTI)

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  8. Autoclave nuclear criticality safety analysis

    SciTech Connect (OSTI)

    D`Aquila, D.M.; Tayloe, R.W. Jr.

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  9. HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

    SciTech Connect (OSTI)

    EVANS, C B

    2004-12-21

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

  10. Supplemental Analysis for the Final Environmental Impact Statement...

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

    Supplemental Analysis for the Final Environmental Impact Statement Supplemental Analysis for the Final Environmental Impact Statement Supplemental Analysis for the Final ...

  11. Fire safety of LPG in marine transportation. Final report

    SciTech Connect (OSTI)

    Martinsen, W.E.; Johnson, D.W.; Welker, J.R.

    1980-06-01

    This report contains an analytical examination of cargo spill and fire hazard potential associated with the marine handling of liquefied petroleum gas (LPG) as cargo. Principal emphasis was on cargo transfer operations for ships unloading at receiving terminals, and barges loading or unloading at a terminal. Major safety systems, including emergency shutdown systems, hazard detection systems, and fire extinguishment and control systems were included in the analysis. Spill probabilities were obtained from fault tree analyses utilizing composite LPG tank ship and barge designs. Failure rates for hardware in the analyses were generally taken from historical data on similar generic classes of hardware, there being very little historical data on the specific items involved. Potential consequences of cargo spills of various sizes are discussed and compared to actual LPG vapor cloud incidents. The usefulness of hazard mitigation systems (particularly dry chemical fire extinguishers and water spray systems) in controlling the hazards posed by LPG spills and spill fires is also discussed. The analysis estimates the probability of fatality for a terminal operator is about 10/sup -6/ to 10/sup -5/ per cargo transfer operation. The probability of fatality for the general public is substantially less.

  12. EA-1212-SA-01: Final Supplement Analysis

    Broader source: Energy.gov [DOE]

    Final Supplement Analysis for the Environmental Assessment for the Lease of Land for the Development of a Research Park at Los Alamos National Laboratory

  13. Final Technical Report on Radioxenon Event Analysis

    SciTech Connect (OSTI)

    Ely, James H.; Cooper, Matthew W.; Hayes, James C.; Heimbigner, Tom R.; McIntyre, Justin I.; Schrom, Brian T.

    2013-03-15

    This is a final deliverable report for the Advanced Spectral Analysis for Radioxenon project with a focus on radioxenon event categorization.

  14. Lawrence Livermore Site Office Safety Basis Self-Assessment Final...

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

    DISCUSSION of RESULTS 1. LSO procedures and mechanisms address and implement 10 CFR 830 ... development and implementation of fully compliant documented safety analyses (DSAs) and ...

  15. Volume II - Accident and Operational Safety Analysis Handbook

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

    208-2012 July 2012 DOE HANDBOOK Accident and Operational Safety Analysis Volume II: ... This Department of Energy (DOE) Accident and Operational Safety Analysis Handbook ...

  16. FAQS Gap Analysis Qualification Card - Nuclear Safety Specialist...

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

    Nuclear Safety Specialist FAQS Gap Analysis Qualification Card - Nuclear Safety Specialist Functional Area Qualification Standard Gap Analysis Qualification Cards outline the ...

  17. FAQS Gap Analysis Qualification Card - Senior Technical Safety...

    Office of Environmental Management (EM)

    Senior Technical Safety Manager FAQS Gap Analysis Qualification Card - Senior Technical Safety Manager Functional Area Qualification Standard Gap Analysis Qualification Cards ...

  18. Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with...

    Office of Scientific and Technical Information (OSTI)

    Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets Citation Details In-Document Search Title: Transient Safety Analysis of Fast Spectrum TRU Burning ...

  19. Waste compatibility safety issues and final results for tank 241-T-110 push mode samples

    SciTech Connect (OSTI)

    Nuzum, J.L.

    1997-05-15

    This document is the final laboratory report for Tank 241-T-110. Push mode core segments were removed from risers 2 and 6 between January 29, 1997, and February 7, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-T-110 Push Mode Core Sampling and analysis Plan (TSAP) and Safety Screening Data Quality Objective (DQO). None of the subsamples submitted for total alpha activity (AT) or differential scanning calorimetry (DSC) analyses exceeded the notification limits stated in DQO.

  20. Nuclear safety for the space exploration initiative. Final report

    SciTech Connect (OSTI)

    Dix, T.E.

    1991-11-01

    The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

  1. Final 2011 Supplement Analysis of the 2005 Final SWEIS | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration | (NNSA) Final 2011 Supplement Analysis of the 2005 Final SWEIS Final 2011 Supplement Analysis of the 2005 Final Site-wide Environmental Impact Statement for Continued Operation of LLNL The below final SUPPLEMENT ANALYSIS is available for public viewing. Final 2011 SUPPLEMENT ANALYSIS of the 2005 Final Site-wide Environmental Impact Statement for Continued Operation of Lawrence Livermore National Laboratory Below are links to the final comments response and final

  2. FP-33 Final Analysis Report

    SciTech Connect (OSTI)

    Williams, R. W.

    2015-10-02

    This report describes an analysis of 244Pu from VNIIEF (FP-33) sent to the Lawrence Livermore National Laboratory as test samples.

  3. DOE's Safety Bulletin No. 2011-01, Events Beyond Design Safety Basis Analysis, March 2011

    Broader source: Energy.gov [DOE]

    PURPOSE This Safety Alert provides information on a safety concern related to the identification and mitigation of events that may fall outside those analyzed in the documented safety analysis.

  4. Fire Risk Implications in Safety Analysis Reports

    SciTech Connect (OSTI)

    Blanchard, A.

    1999-03-31

    Fire can be a significant risk for facilities that store and handle radiological material. Such events must be evaluated as part of a comprehensive safety analysis. SRS has been developing methods to evaluate radiological fire risk in such facilities. These methods combined with the analysis techniques proposed by DOE-STD-3009-94 have provided a better understanding of how fire risks in nuclear facilities should be managed. To ensure that these new insights are properly disseminated the DOE Savannah River Office and the Defense Nuclear Facility Safety Board (DNFSB) requested Westinghouse Savannah River Company (WSRC) prepare this paper.

  5. Technical Standards, Safety Analysis Toolbox Codes - November 2003 |

    Office of Environmental Management (EM)

    Department of Energy Standards, Safety Analysis Toolbox Codes - November 2003 Technical Standards, Safety Analysis Toolbox Codes - November 2003 November 2003 Software Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes Safety analysis software for the DOE "toolbox" was designated by DOE/EH in March 2003 (DOE/EH, 2003). The supporting basis for this designation was provided by a DOE-chartered Safety Analysis Software Group in the technical report, Selection of

  6. K West integrated water treatment system subproject safety analysis document

    SciTech Connect (OSTI)

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

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

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

    December 2, 2014 New Nuclear Facility Documented Safety Analysis and Technical Safety Requirements Criteria Review and Approach Document (EA CRAD 31-07, Rev. 0) CRAD, New Nuclear...

  8. SYNTHESIS OF SAFETY ANALYSIS AND FIRE HAZARD ANALYSIS METHODOLOGIES

    SciTech Connect (OSTI)

    Coutts, D

    2007-04-17

    Successful implementation of both the nuclear safety program and fire protection program is best accomplished using a coordinated process that relies on sound technical approaches. When systematically prepared, the documented safety analysis (DSA) and fire hazard analysis (FHA) can present a consistent technical basis that streamlines implementation. If not coordinated, the DSA and FHA can present inconsistent conclusions, which can create unnecessary confusion and can promulgate a negative safety perception. This paper will compare the scope, purpose, and analysis techniques for DSAs and FHAs. It will also consolidate several lessons-learned papers on this topic, which were prepared in the 1990s.

  9. Hanford safety analysis and risk assessment handbook (SARAH)

    SciTech Connect (OSTI)

    GARVIN, L.J.

    2003-01-20

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 1,2, and 3 U.S. Department of Energy (DOE) nuclear facilities. SARAH describes currently acceptable methodology for development of a Documented Safety Analysis (DSA) and derivation of technical safety requirements (TSR) based on 10 CFR 830, ''Nuclear Safety Management,'' Subpart B, ''Safety Basis Requirements,'' and provides data to ensure consistency in approach.

  10. 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

  11. Safety analysis report for packaging upgrade plan

    SciTech Connect (OSTI)

    Kelly, D.L., Westinghouse Hanford

    1996-12-09

    This Safety Analysis Report for Packaging (SARP) Upgrade Plan reflects a SARP upgrade schedule based on the most current program needs. A performance agreement has been assigned, beginning in FY 1997, to update, revise, and/or cancel 20 percent of the existing onsite SARPS, so that 100 percent are reviewed and within current standards by the completion of the Project Hanford Management Contract (five-year period).

  12. LPG land transportation and storage safety. Final report

    SciTech Connect (OSTI)

    Not Available

    1981-09-01

    This report contains an analytical examination of fatal accidents involving liquefied petroleum gas (LPG) releases during transportation and/or transportation related storage. Principal emphasis was on accidents during the nine-year period 1971 to 1979. Fatalities to members of the general public (i.e., those at the scene of the accident through coincidence or curiosity) were of special interest. Transportation accidents involving railroad tank cars, trucks, and pipelines were examined as were accidents at storage facilities, including loading and unloading at such facilities. The main sources of the necessary historical accident data were the accident reports submitted to the Department of Transportation by LPG carriers, National Transportation Safety Board accident reports, articles in the National Fire Protection Association journals, other literature, and personal interviews with firemen, company personnel, and others with knowledge of certain accidents. The data indicate that, on the average, releases of LPG during transportation and intermediate storage cause approximately six fatalities per year to members of the general public. The individual risk is about 1 death per 37,000,000 persons; about the same as the risk of a person on the ground being killed by an airplane crash, and much less than the risk of death by lightning, tornadoes, or dam failures.

  13. LPG land transportation and storage safety. Final report

    SciTech Connect (OSTI)

    Martinsen, W.E.; Cavin, W.D.

    1981-09-01

    This report contains an analytical examination of fatal accidents involving liquefied petroleum gas (LPG) releases during transportation and/or transportation related storage. Principal emphasis was on accidents during the nine-year period 1971 through 1979. Fatalities to members of the general public (i.e., those at the scene of the accident through coincidence or curiosity) were of special interest. Transportation accidents involving railroad tank cars, trucks, and pipelines were examined as were accidents at storage facilities, including loading and unloading at such facilities. The main sources of the necessary historical accident data were the accident reports submitted to the Department of Transportation by LPG carriers, National Transportation Safety Board accident reports, articles in the National Fire Protection Association journals, other literature, and personal interviews with firemen, company personnel, and others with knowledge of certain accidents. The data indicate that, on the average, releases of LPG during transportation and intermediate storage cause approximately six fatalities per year to members of the general public. The individual risk is about 1 death per 37,000,000 persons; about the same as the risk of a person on the ground being killed by an airplane crash, and much less than the risk of death by lightning, tornadoes, or dam failures.

  14. EIS-0350-SA-02: Final Supplement Analysis | Department of Energy

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

    350-SA-02: Final Supplement Analysis EIS-0350-SA-02: Final Supplement Analysis Chemistry and Metallurgy Research Building Replacement Project at Los Alamos National Laboratory, Los...

  15. Office of Environmental Protection, Sustainability Support & Corporate Safety Analysis

    Broader source: Energy.gov [DOE]

    Lists the mission and functions of the Office of Environmental Protection, Sustainability Support & Corporate Safety Analysis (SESA)

  16. Final Report Computational Analysis of Dynamical Systems

    SciTech Connect (OSTI)

    Guckenheimer, John

    2012-05-08

    This is the final report for DOE Grant DE-FG02-93ER25164, initiated in 1993. This grant supported research of John Guckenheimer on computational analysis of dynamical systems. During that period, seventeen individuals received PhD degrees under the supervision of Guckenheimer and over fifty publications related to the grant were produced. This document contains copies of these publications.

  17. Monthly Analysis of Electrical Safety Occurrences - January 2011...

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

    information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - January 2011 More Documents & Publications Monthly Analysis of...

  18. Monthly Analysis of Electrical Safety Occurrences - May 2011...

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

    information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - May 2011 More Documents & Publications Monthly Analysis of...

  19. Monthly Analysis of Electrical Safety Occurrences - April 2011...

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

    information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - April 2011 More Documents & Publications Monthly Analysis of...

  20. Monthly Analysis of Electrical Safety Occurrences - March 2011...

    Energy Savers [EERE]

    analysis of March 2011 and past information for information exchange and continual learning. PDF icon Monthly Analysis of Electrical Safety Occurrences - March 2011 More...

  1. Safety System Oversight Staffing Analysis - Blank Sheet | Department...

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

    Safety System Oversight Staffing Analysis - Blank Sheet This Staffing Analysis calculation is completed using an Excel worksheet. Information locations are identified by titles in ...

  2. Safety analysis report for packaging upgrade plan

    SciTech Connect (OSTI)

    KELLY, D.L.

    1998-11-18

    This Safety Analysis Report for Packaging (SARP) Upgrade Plan reflects a revised SARP upgrade schedule based on the most current program needs. A Project Hanford Management Contract (PHMC) Performance Expectation exists to update, revise, and/or cancel seven onsite SARPS during FY 1999. It is the U.S. Department of Energy's desire that 100% of the SARPs (which existed at the beginning of the PHMC Contract) be upgraded, revised, and/or canceled by the end of the five year contract. This plan is a ''living'' document and is used as a management tool.

  3. 242-A evaporator safety analysis report

    SciTech Connect (OSTI)

    CAMPBELL, T.A.

    1999-05-17

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR.

  4. Regulatory analysis technical evaluation handbook. Final report

    SciTech Connect (OSTI)

    1997-01-01

    The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of quality regulatory analysis documents and to implement the policies of the Regulatory Analysis Guidelines of the US Nuclear Regulatory Commission (NUREG/BR-0058 Rev. 2). This Handbook expands upon policy concepts included in the NRC Guidelines and translates the six steps in preparing regulatory analyses into implementable methodologies for the analyst. It provides standardized methods of preparation and presentation of regulatory analyses, with the inclusion of input that will satisfy all backfit requirements and requirements of NRC`s Committee to Review Generic Requirements. Information on the objectives of the safety goal evaluation process and potential data sources for preparing a safety goal evaluation is also included. Consistent application of the methods provided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions. The handbook is being issued in loose-leaf format to facilitate revisions. NRC intends to periodically revise the handbook as new and improved guidance, data, and methods become available.

  5. Monthly Analysis of Electrical Safety Occurrences - February 2013 |

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

    Department of Energy 3 Monthly Analysis of Electrical Safety Occurrences - February 2013 February 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of February 2013 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - February 2013 (2.74 MB)

  6. Monthly Analysis of Electrical Safety Occurrences - July 2012 |

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

    Department of Energy 2 Monthly Analysis of Electrical Safety Occurrences - July 2012 July 2012 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of July 2012 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - July 2012 (1.71 MB) More Documents

  7. Monthly Analysis of Electrical Safety Occurrences - July 2013 |

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

    Department of Energy 3 Monthly Analysis of Electrical Safety Occurrences - July 2013 July 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of July 2013 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - July 2013 (2.69 MB) More Documents

  8. Monthly Analysis of Electrical Safety Occurrences - March 2013 |

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

    Department of Energy 3 Monthly Analysis of Electrical Safety Occurrences - March 2013 March 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of March 2013 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - March 2013 (2.61 MB) More Documents

  9. Monthly Analysis of Electrical Safety Occurrences - November 2012 |

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

    Department of Energy 2 Monthly Analysis of Electrical Safety Occurrences - November 2012 November 2012 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of November 2012 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - November 2012 (2.17 MB)

  10. Monthly Analysis of Electrical Safety Occurrences - October 2012 |

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

    Department of Energy 2 Monthly Analysis of Electrical Safety Occurrences - October 2012 October 2012 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of October 2012 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - October 2012 (1.52 MB) More

  11. Monthly Analysis of Electrical Safety Occurrences - September 2011 |

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

    Department of Energy 1 Monthly Analysis of Electrical Safety Occurrences - September 2011 September 2011 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of September 2011 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - September 2011 (679.94

  12. Monthly Analysis of Electrical Safety Occurrences - September 2013 |

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

    Department of Energy 3 Monthly Analysis of Electrical Safety Occurrences - September 2013 September 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning. The report discusses the quantitative analysis of September 2013 and past information for information exchange and continual learning. Monthly Analysis of Electrical Safety Occurrences - September 2013 (3.6 MB)

  13. CRAD, Documented Safety Analysis Development- April 23, 2013

    Broader source: Energy.gov [DOE]

    Review of Documented Safety Analysis Development for the Hanford Site Waste Treatment and Immobilization Plant (LBL Facilities) (HSS CRAD 45-58, Rev. 0)

  14. Monthly Analysis of Electrical Safety Occurrences – April 2013

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  15. Safety analysis of in-use vehicle wrapping cylinder

    Broader source: Energy.gov [DOE]

    The focus of this presentation is on the security analysis for wrapped cylinders used in vehicles and analyzing safety conditions and environmental effects through testing.

  16. Monthly Analysis of Electrical Safety Occurrences – December 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  17. Monthly Analysis of Electrical Safety Occurrences – March 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  18. Monthly Analysis of Electrical Safety Occurrences – September 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  19. Monthly Analysis of Electrical Safety Occurrences – July 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  20. Monthly Analysis of Electrical Safety Occurrences – June 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  1. Monthly Analysis of Electrical Safety Occurrences – June 2013

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  2. Monthly Analysis of Electrical Safety Occurrences – February 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  3. Monthly Analysis of Electrical Safety Occurrences – October 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  4. Monthly Analysis of Electrical Safety Occurrences – May 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  5. Monthly Analysis of Electrical Safety Occurrences – August 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  6. Monthly Analysis of Electrical Safety Occurrences – June 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  7. Monthly Analysis of Electrical Safety Occurrences – August 2013

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  8. Monthly Analysis of Electrical Safety Occurrences – November 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  9. Monthly Analysis of Electrical Safety Occurrences – April 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  10. Monthly Analysis of Electrical Safety Occurrences – January 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  11. Monthly Analysis of Electrical Safety Occurrences – August 2011

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  12. Monthly Analysis of Electrical Safety Occurrences – May 2013

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  13. Monthly Analysis of Electrical Safety Occurrences – January 2013

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  14. Monthly Analysis of Electrical Safety Occurrences – December 2012

    Broader source: Energy.gov [DOE]

    An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and continual learning.

  15. 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.

  16. Additional EIPC Study Analysis. Final Report

    SciTech Connect (OSTI)

    Hadley, Stanton W; Gotham, Douglas J.; Luciani, Ralph L.

    2014-12-01

    Between 2010 and 2012 the Eastern Interconnection Planning Collaborative (EIPC) conducted a major long-term resource and transmission study of the Eastern Interconnection (EI). With guidance from a Stakeholder Steering Committee (SSC) that included representatives from the Eastern Interconnection States Planning Council (EISPC) among others, the project was conducted in two phases. Phase 1 involved a long-term capacity expansion analysis that involved creation of eight major futures plus 72 sensitivities. Three scenarios were selected for more extensive transmission- focused evaluation in Phase 2. Five power flow analyses, nine production cost model runs (including six sensitivities), and three capital cost estimations were developed during this second phase. The results from Phase 1 and 2 provided a wealth of data that could be examined further to address energy-related questions. A list of 14 topics was developed for further analysis. This paper brings together the earlier interim reports of the first 13 topics plus one additional topic into a single final report.

  17. Lng vehicle technology, economics, and safety assessment. Final report, April 1991-June 1993

    SciTech Connect (OSTI)

    Powars, C.A.; Moyer, C.B.; Lowell, D.D.

    1994-02-01

    Liquid natural gas (LNG) is an attractive transportation fuel because of its high heating value and energy density (i.e. Btu/lb and Btu/gal), clean burning characteristics, relatively low cost ($/Btu), and domestic availability. This research evaluated LNG vehicle and refueling system technology, economics, and safety. Prior and current LNG vehicle projects were studied to identify needed technology improvements. Life-cycle cost analyses considered various LNG vehicle and fuel supply options. Safety records, standards, and analysis methods were reviewed. The LNG market niche is centrally fueled heavy-duty fleet vehicles with high fuel consumption. For these applications, fuel cost savings can amortize equipment capital costs.

  18. Safety analysis report for the Waste Storage Facility. Revision 2

    SciTech Connect (OSTI)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  19. SNF fuel retrieval sub project safety analysis document

    SciTech Connect (OSTI)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  20. Final Review of Safety Assessment Issues at Savannah River Site, August 2011

    SciTech Connect (OSTI)

    Napier, Bruce A.; Rishel, Jeremy P.; Bixler, Nathan E.

    2011-12-15

    At the request of Savannah River Nuclear Solutions (SRNS) management, a review team composed of experts in atmospheric transport modeling for environmental radiation dose assessment convened at the Savannah River Site (SRS) on August 29-30, 2011. Though the meeting was prompted initially by suspected issues related to the treatment of surface roughness inherent in the SRS meteorological dataset and its treatment in the MELCOR Accident Consequence Code System Version 2 (MACCS2), various topical areas were discussed that are relevant to performing safety assessments at SRS; this final report addresses these topical areas.

  1. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect (OSTI)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  2. Safety Analysis Report Update Program: Overview and Phase 1 implementation

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

    During FY 1989, the DOE-Oak Ridge Operations (ORO) office formed joint Operating Contractor/DOE-ORO organizations to address safety analysis related issues. The Safety Analysis Report Working Group (SARWG) took on the task of developing a strategy to address the issue of updating SARs to today's standards. The resulting SAR Update Program was approved by the Safety Analysis Report Management Group (SARMG) and on November 6, 1989, was accepted by the senior management of DOE-ORO, and its operating contractors, including Martin Marietta Energy Systems, Inc. This SAR Update Program consists of five phases: Phase 0 -- continued operation evaluations; Phase 1 -- hazard classification and qualitative analysis; Phase 1A -- updated operational safety requirements; Phase 2 -- quantitative accident analysis; and, Phase 3 -- complete DOE-approved SARs. 8 refs., 17 figs., 6 tabs.

  3. Planning Document for an NBSR Conversion Safety Analysis Report

    SciTech Connect (OSTI)

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

    2013-09-25

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted.This report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

  4. CRAD, Preliminary Documented Safety Analysis - July 25, 2014...

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

    4 Preliminary Documented Safety Analysis (IEA CRAD 31-2, REV. 0) This Criteria Review and Approach Document (IEA CRAD 31-2, REV. 0) provides objectives, criteria, and approaches...

  5. Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis

    Broader source: Energy.gov [DOE]

    This paper addresses why the use of an Integrated Safety Analysis (“ISA”) is appropriate for fuel recycling facilities1 which would be licensed under new regulations currently being considered by...

  6. Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis

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

    2014-11-12

    This Department of Energy (DOE) Standard (STD), DOE-STD-3009-2014, describes a method for preparing a Documented Safety Analysis (DSA) that is acceptable to DOE for nonreactor nuclear facilities.

  7. ITP Glass: Industrial Glass Bandwidth Analysis Final Report,...

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

    Wolf Gas Technology Institute Energy Utilization Center August 2007 Industrial Glass Bandwidth Analysis FINAL REPORT Prepared by: David M. Rue James Servaites Dr. Warren Wolf ...

  8. Safety analysis approaches or mixed transuranic waste.

    SciTech Connect (OSTI)

    Courtney, J. C.; Dwight, C. C.; Forrester, R. J.; Lehto, M. A.; Pan, Y. C.

    1999-02-10

    Argonne National Laboratory (ANL) has completed a survey of assumptions and techniques used for safety analyses at seven sites that handle or store mixed transuranic (TRU) waste operated by contractors for the US Department of Energy (DOE). While approaches to estimating on-site and off-site consequences of hypothetical accidents differ, there are commonalities in all of the safety studies. This paper identifies key parameters and methods used to estimate the radiological consequences associated with release of waste forms under abnormal conditions. Specific facilities are identified by letters with their safety studies listed in a bibliography rather than as specific references so that similarities and differences are emphasized in a nonjudgmental manner. References are provided for specific parameters used to project consequences associated with compromise of barriers and dispersion of potentially hazardous materials. For all of the accidents and sites, estimated dose commitments are well below guidelines even using highly conservative assumptions. Some of the studies quantified the airborne concentrations of toxic materials; this paper only addresses these analyses briefly, as an entire paper could be dedicated to this subject.

  9. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    SciTech Connect (OSTI)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  10. The conservation planning analysis model. Final report

    SciTech Connect (OSTI)

    Not Available

    1993-12-31

    This paper contains the source code for a program on conservation planning analysis for residential, commercial and industrial customers.

  11. Safety analysis report for packaging (onsite) steel drum

    SciTech Connect (OSTI)

    McCormick, W.A.

    1998-09-29

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum.

  12. Final Report- Hydrogen Delivery Infrastructure Options Analysis

    Office of Energy Efficiency and Renewable Energy (EERE)

    This report provides in-depth analysis of various hydrogen delivery options to determine the most cost effective infrastructure and R&D efforts for the long term.

  13. EIS-0350-SA-02: Final Supplement Analysis | Department of Energy

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

    50-SA-02: Final Supplement Analysis EIS-0350-SA-02: Final Supplement Analysis Chemistry and Metallurgy Research Building Replacement Project at Los Alamos National Laboratory, Los Alamos, New Mexico Supplement Analysis To meet mission requirements, the U.S. Department of Energy/National Nuclear Security Administration (DOE/NNSA) proposes to relocate analytical chemistry (AC) and materials characterization (MC) capabilities at Los Alamos National Laboratory (LANL) from the Chemistry and

  14. Safety

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

    safety Safety All JLF participants must comply fully with all LLNL safety regulations and procedures by becoming a Registered User of the facility. All JLF participants must complete available LLNL safety training: HS5200-W Laser Safety HS4258-W Beryllium Awareness HS4261-W Lead Awareness HS5220-W Electrical Safety Awareness HS6001-W General Employee Radiological HS4240-W Chemical Safety HS4680-W PPE To access these training modules link here [LTRAIN] from inside LLNL, or here from anywhere. All

  15. Final characterization and safety screen report of double shell tank 241-AP-105 for evaporator campaign 97-1

    SciTech Connect (OSTI)

    Miller, G.L.

    1997-01-20

    Evaporator candidate feed from tank 241-AP-105 (hereafter referred to as AP-105) was characterized for physical, inorganic, organic and radiochemical parameters by the 222-S Laboratory as directed by the Tank Sample and Analysis Plan (TSAP), References 1 through 4, and Engineering Change Notice, number 635332, Reference 5. This data package satisfies the requirement for a format IV, final report as described in Reference 1. This data package is also a follow-up to the 45-Day safety screen results for tank AP-105, Reference 8, which was issued on November 5, 1996, and is attached as Section II to this report. Preliminary data in the form of summary analytical tables were provided to the project in advance of this final report to enable early estimation of evaporator operational parameters, using the Predict modeling program. Analyses were performed at the 222-S Laboratory as defined and specified in the TSAP and the Laboratory's Quality Assurance P1an, References 6 and 7. Any deviations from the instructions documented in the TSAP are discussed in this narrative and are supported with additional documentation.

  16. Corporate Analysis of DOE Safety Performance | Department of...

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

    Current Safety Performance Trends The Office of Environment, Health, Safety and Security, ... Strategic Safety Goals Occupational Safety Performance Occurrence Reporting Trends For ...

  17. Analysis of Hybrid Hydrogen Systems: Final Report

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

    U.S. Energy Information Administration Analysis of Heat Rate Improvement Potential at Coal-Fired Power Plants Release date: May 19, 2015 Introduction The thermal efficiency of electricity production is represented by the heat rate, which measures the amount of energy used to generate one kilowatthour of electricity.1 A generating unit with a lower, or more efficient, heat rate can generate the same quantity of electricity while consuming less fuel, compared with a unit with a higher heat

  18. ITP Glass: Industrial Glass Bandwidth Analysis Final Report, August 2007 |

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

    Department of Energy Industrial Glass Bandwidth Analysis Final Report, August 2007 ITP Glass: Industrial Glass Bandwidth Analysis Final Report, August 2007 industrial_bandwidth.pdf (360.99 KB) More Documents & Publications ITP Glass: Glass Industry of the Future: Energy and Environmental Profile of the U.S. Glass Industry; April, 2002 ITP Glass: Glass Industry Technology Roadmap; April 2002 ITP Glass: A Clear Vision for a Bright Future

  19. Final report on the Pathway Analysis Task

    SciTech Connect (OSTI)

    Whicker, F.W.; Kirchner, T.B.

    1993-04-01

    The Pathway Analysis Task constituted one of several multi-laboratory efforts to estimate radiation doses to people, considering all important pathways of exposure, from the testing of nuclear devices at the Nevada Test Site (NTS). The primary goal of the Pathway Analysis Task was to predict radionuclide ingestion by residents of Utah, Nevada, and portions of seven other adjoining western states following radioactive fallout deposition from individual events at the NTS. This report provides comprehensive documentation of the activities and accomplishments of Colorado State University`s Pathway Analysis Task during the entire period of support (1979--91). The history of the project will be summarized, indicating the principal dates and milestones, personnel involved, subcontractors, and budget information. Accomplishments, both primary and auxiliary, will be summarized with general results rather than technical details being emphasized. This will also serve as a guide to the reports and open literature publications produced, where the methodological details and specific results are documented. Selected examples of results on internal dose estimates are provided in this report because the data have not been published elsewhere.

  20. Implementing 10 CFR 830 at the FEMP Silos: Nuclear Health and Safety Plans as Documented Safety Analysis

    SciTech Connect (OSTI)

    Fisk, Patricia; Rutherford, Lavon

    2003-06-01

    The objective of the Silos Project at the Fernald Closure Project (FCP) is to safely remediate high-grade uranium ore residues (Silos 1 and 2) and metal oxide residues (Silo 3). The evolution of Documented Safety Analyses (DSAs) for these facilities has reflected the changes in remediation processes. The final stage in silos DSAs is an interpretation of 10 CFR 830 Safe Harbor Requirements that combines a Health and Safety Plan with nuclear safety requirements. This paper will address the development of a Nuclear Health and Safety Plan, or N-HASP.

  1. Safety analysis of optically ignited explosive and pyrotechnic devices

    SciTech Connect (OSTI)

    Merson, J.A.; Salas, F.J.; Holswade, S.

    1994-05-01

    The future of optical ordnance depends on the acceptance, validation and verification of the stated safety enhancement claims of optical ordnance over existing electrical explosive devices (EED`s). Sandia has been pursuing the development of optical ordnance, with the primary motivation of this effort being the enhancement of explosive safety by specifically reducing the potential of premature detonation that can occur with low energy electrically ignited explosive devices. By using semiconductor laser diodes for igniting these devices, safety improvements can be made without being detrimental to current system concerns since the inputs required for these devices are similar to electrical systems. Laser Diode Ignition (LDI) of the energetic material provides the opportunity to remove the bridgewire and electrically conductive pins from the charge cavity, creating a Faraday cage and thus isolating the explosive or pyrotechnic materials from stray electrical ignition sources. Recent results from our continued study of safety enhancements are presented. The areas of investigation which are presented include: (1) unintended optical source analysis, specifically lightning insensitivity, (2) electromagnetic radiation (EMR) and electrostatic discharge (ESD) insensitivity analysis, and (3) powder safety.

  2. Hazard Analysis Database report

    SciTech Connect (OSTI)

    Niemi, B.J.

    1997-08-12

    This document describes and defines the Hazard Analysis Database for the Tank Waste Remediation System Final Safety Analysis Report.

  3. PAT-1 safety analysis report addendum.

    SciTech Connect (OSTI)

    Weiner, Ruth F.; Schmale, David T.; Kalan, Robert J.; Akin, Lili A.; Miller, David Russell; Knorovsky, Gerald Albert; Yoshimura, Richard Hiroyuki; Lopez, Carlos; Harding, David Cameron; Jones, Perry L.; Morrow, Charles W.

    2010-09-01

    The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging with the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.

  4. Safety analysis report for packaging (onsite) multicanister overpack cask

    SciTech Connect (OSTI)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  5. Worker Safety and Health and Nuclear Safety Quarterly Performance Analysis (January - March 2008)

    SciTech Connect (OSTI)

    Kerr, C E

    2009-10-07

    The DOE Office of Enforcement expects LLNL to 'implement comprehensive management and independent assessments that are effective in identifying deficiencies and broader problems in safety and security programs, as well as opportunities for continuous improvement within the organization' and to 'regularly perform assessments to evaluate implementation of the contractor's processes for screening and internal reporting.' LLNL has a self-assessment program, described in ES&H Manual Document 4.1, that includes line, management and independent assessments. LLNL also has in place a process to identify and report deficiencies of nuclear, worker safety and health and security requirements. In addition, the DOE Office of Enforcement expects LLNL to evaluate 'issues management databases to identify adverse trends, dominant problem areas, and potential repetitive events or conditions' (page 14, DOE Enforcement Process Overview, December 2007). LLNL requires that all worker safety and health and nuclear safety noncompliances be tracked as 'deficiencies' in the LLNL Issues Tracking System (ITS). Data from the ITS are analyzed for worker safety and health (WSH) and nuclear safety noncompliances that may meet the threshold for reporting to the DOE Noncompliance Tracking System (NTS). This report meets the expectations defined by the DOE Office of Enforcement to review the assessments conducted by LLNL, analyze the issues and noncompliances found in these assessments, and evaluate the data in the ITS database to identify adverse trends, dominant problem areas, and potential repetitive events or conditions. The report attempts to answer three questions: (1) Is LLNL evaluating its programs and state of compliance? (2) What is LLNL finding? (3) Is LLNL appropriately managing what it finds? The analysis in this report focuses on data from the first quarter of 2008 (January through March). This quarter is analyzed within the context of information identified in previous quarters to

  6. Microsoft PowerPoint - TrafficSafetyBriefing FINAL.pptx

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

    of the Independent Results of the Independent Hanford Traffic Safety Hanford Traffic Safety Hanford Traffic Safety Hanford Traffic Safety Study and Next Steps Study and Next Steps May 2010 Background Background * Your safety does not stop with the work you do - driving safely on site is an equal factor in keeping you g y q p g y safe. * Key Hanford roadways are operating over capacity at k ti i ti d l d peak times, causing congestion and slow-downs. * In some spots the posted speed limits do not

  7. Safety

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

    Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering ...

  8. safety

    National Nuclear Security Administration (NNSA)

    contractor at the Nevada National Security Site, has been recognized by the Department of Energy for excellence in occupational safety and health protection. National Nuclear...

  9. Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

    SciTech Connect (OSTI)

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

  10. Improving the regulation of safety at DOE nuclear facilities. Final report

    SciTech Connect (OSTI)

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

  11. CRAD, New Nuclear Facility Documented Safety Analysis and Technical Safety Requirements- December 2, 2014 (EA CRAD 31-07, Rev. 0)

    Broader source: Energy.gov [DOE]

    New Nuclear Facility Documented Safety Analysis and Technical Safety Requirements Criteria Review and Approach Document (EA CRAD 31-07, Rev. 0)

  12. Methods and criteria for safety analysis (FIN L2535)

    SciTech Connect (OSTI)

    Not Available

    1992-12-01

    In response to the NRC request for a proposal dated October 20, 1992, Westinghouse Savannah River Company (WSRC) submit this proposal to provide contractural assistance for FIN L2535, ``Methods and Criteria for Safety Analysis,`` as specified in the Statement of Work attached to the request for proposal. The Statement of Work involves development of safety analysis guidance for NRC licensees, arranging a workshop on this guidance, and revising NRC Regulatory Guide 3.52. This response to the request for proposal offers for consideration the following advantages of WSRC in performing this work: Experience, Qualification of Personnel and Resource Commitment, Technical and Organizational Approach, Mobilization Plan, Key Personnel and Resumes. In addition, attached are the following items required by the NRC: Schedule II, Savannah River Site - Job Cost Estimate, NRC Form 189, Project and Budget Proposal for NRC Work, page 1, NRC Form 189, Project and Budget Proposal for NRC Work, page 2, Project Description.

  13. Safety analysis report for packaging (onsite) sample pig transport system

    SciTech Connect (OSTI)

    MCCOY, J.C.

    1999-03-16

    This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document.

  14. ORNL Evaluation of Electrabel Safety Cases for Doel 3 / Tihange 2: Final Report

    SciTech Connect (OSTI)

    Bass, Bennett Richard; Dickson, Terry L.; Gorti, Sarma B.; Klasky, Hilda B.; Nanstad, Randy K.; Sokolov, Mikhail A.; Williams, Paul T.; Server, W. L.

    2015-11-01

    Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasilaminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, New York (1992 and 2004). That screening assessment of all EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, Alternative Characterization Rules for QuasiLaminar Flaws . Results and conclusions from the ORNL flaw acceptance assessments of D3/T2 were compared with those from the 2015 EBL Safety Cases. Specific findings of the ORNL evaluation of that part of the EBL structural integrity assessment focusing on stability of the flaw population subjected to primary design transients include the following: ORNL s analysis results were similar to those of EBL in that very few characterized flaws were found not compliant with the ASME (1992) acceptance criterion. ORNL s application of the more recent ASME Section XI (2004) produced only four noncompliant flaws, all due to LOCAs. The finding of a greater number of non-compliant flaws in the EBL screening assessment is due principally to a significantly more restrictive (conservative) criterion for flaw size acceptance used by EBL. ORNL s screening assessment results

  15. Guidance for preparation of safety analysis reports for nonreactor facilities and operations. Revision 2

    SciTech Connect (OSTI)

    Not Available

    1992-11-06

    Department of Energy (DOE) Orders 5480.23, ``Nuclear Safety Analysis Reports,`` and 5481.1B, ``Safety Analysis and Review System`` require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports.

  16. Guidance for preparation of safety analysis reports for nonreactor facilities and operations

    SciTech Connect (OSTI)

    Not Available

    1992-11-06

    Department of Energy (DOE) Orders 5480.23, Nuclear Safety Analysis Reports,'' and 5481.1B, Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports.

  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. A ''Toolbox''21 Equivalent Process for Safety Analysis Software

    SciTech Connect (OSTI)

    O'KULA, KR

    2004-04-30

    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 that prevent or mitigate potential accidents. The development and maintenance of a collection, or ''toolbox,'' of multiple-site use, standard solution, Software Quality Assurance (SQA)-compliant safety software is one of the major improvements identified in the associated DOE Implementation Plan (IP). The DOE safety analysis toolbox will contain a set of appropriately quality-assured, configuration-controlled, safety analysis codes, recognized for DOE-broad, safety basis applications. Currently, six widely applied safety analysis computer codes have been designated for toolbox consideration. While the toolbox concept considerably reduces SQA burdens among DOE users of these codes, many users of unique, single-purpose, or single-site software may still have sufficient technical justification to continue use of their computer code of choice, but are thwarted by the multiple-site condition on toolbox candidate software. The process discussed here provides a roadmap for an equivalency argument, i.e., establishing satisfactory SQA credentials for single-site software that can be deemed ''toolbox-equivalent''. The process is based on the model established to meet IP Commitment 4.2.1.2: Establish SQA criteria for the safety analysis ''toolbox'' codes. Implementing criteria that establish the set of prescriptive SQA requirements are based on implementation plan/procedures from the Savannah River Site, also incorporating aspects of those from the Waste Isolation Pilot Plant (SNL component) and the Yucca Mountain Project. The major requirements are met with evidence of a software quality assurance plan, software requirements and design documentation, user's instructions, test report, a

  19. Management of radioactive material safety programs at medical facilities. Final report

    SciTech Connect (OSTI)

    Camper, L.W.; Schlueter, J.; Woods, S.

    1997-05-01

    A Task Force, comprising eight US Nuclear Regulatory Commission and two Agreement State program staff members, developed the guidance contained in this report. This report describes a systematic approach for effectively managing radiation safety programs at medical facilities. This is accomplished by defining and emphasizing the roles of an institution`s executive management, radiation safety committee, and radiation safety officer. Various aspects of program management are discussed and guidance is offered on selecting the radiation safety officer, determining adequate resources for the program, using such contractual services as consultants and service companies, conducting audits, and establishing the roles of authorized users and supervised individuals; NRC`s reporting and notification requirements are discussed, and a general description is given of how NRC`s licensing, inspection and enforcement programs work.

  20. FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE

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

    HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE November 14, 2013 Richland, WA Topics in this Meeting Summary Opening ......................................................................................................................................................... 1 DOE's Emergency Preparedness Program - Joint with the River and Plateau (RAP) Committee and Public Information and Communication (PIC) Committee ..........................................................................

  1. DOE's Approach to Nuclear Facility Safety Analysis and Management

    Broader source: Energy.gov [DOE]

    Presenter: Dr. James O'Brien, Director, Office of Nuclear Safety, Office of Health, Safety and Security, US Department of Energy

  2. NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21

    SciTech Connect (OSTI)

    Smith, G L

    1989-12-01

    The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.

  3. Application of coupled codes for safety analysis and licensing issues

    SciTech Connect (OSTI)

    Langenbuch, S.; Velkov, K.

    2006-07-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  4. Review guidelines on software languages for use in nuclear power plant safety systems. Final report

    SciTech Connect (OSTI)

    Hecht, H.; Hecht, M.; Graff, S.; Green, W.; Lin, D.; Koch, S.; Tai, A.; Wendelboe, D.

    1996-06-01

    Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada, C/C++, Programmable Logic Controller (PLC) Ladder Logic, International Electrotechnical Commission (IEC) Standard 1131-3 Sequential Function Charts, Pascal, and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.s

  5. Commercial Vehicle Safety Alliance (CVSA)/Department of Energy (DOE) cooperative agreement final report

    SciTech Connect (OSTI)

    Slavich, Antoinette; Daust, James E.

    1999-10-01

    This S and T product is a culmination of the activities, including research of the Commercial Vehicle Safety Alliance (CVSA) in developing and implementing inspection procedures and the out-of-service criteria for states and tribes to use when inspecting HRCQ and Transuranic shipments of radioactive materials. The report also contains the results of a pilot study to test the procedures.

  6. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    SciTech Connect (OSTI)

    Herborn, D.I.

    1992-10-01

    This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals for the Hanford Waste Vitrification Plant.

  7. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    SciTech Connect (OSTI)

    Not Available

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

  8. Canister storage building (CSB) safety analysis report phase 3: Safety analysis documentation supporting CSB construction

    SciTech Connect (OSTI)

    Garvin, L.J.

    1997-04-28

    The Canister Storage Building (CSB) will be constructed in the 200 East Area of the U.S. Department of Energy (DOE) Hanford Site. The CSB will be used to stage and store spent nuclear fuel (SNF) removed from the Hanford Site K Basins. The objective of this chapter is to describe the characteristics of the site on which the CSB will be located. This description will support the hazard analysis and accident analyses in Chapter 3.0. The purpose of this report is to provide an evaluation of the CSB design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB.

  9. Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

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

    Di Maio, Francesco; Zio, Enrico; Smith, Curtis; Rychkov, Valentin

    2015-07-06

    The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs andmore » activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

  10. Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

    SciTech Connect (OSTI)

    Di Maio, Francesco; Zio, Enrico; Smith, Curtis; Rychkov, Valentin

    2015-07-06

    The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs and activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).

  11. Documented Safety Analysis for the B695 Segment

    SciTech Connect (OSTI)

    Laycak, D

    2008-09-11

    This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, or {sup 3}H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building systems

  12. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-06-20

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

  13. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect (OSTI)

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 keff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  14. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect (OSTI)

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  15. Documented Safety Analysis for the Waste Storage Facilities

    SciTech Connect (OSTI)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  16. Documented Safety Analysis for the Waste Storage Facilities March 2010

    SciTech Connect (OSTI)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  17. Hazard screening application guide. Safety Analysis Report Update Program

    SciTech Connect (OSTI)

    1992-06-01

    The basic purpose of hazard screening is to group precesses, facilities, and proposed modifications according to the magnitude of their hazards so as to determine the need for and extent of follow on safety analysis. A hazard is defined as a material, energy source, or operation that has the potential to cause injury or illness in human beings. The purpose of this document is to give guidance and provide standard methods for performing hazard screening. Hazard screening is applied to new and existing facilities and processes as well as to proposed modifications to existing facilities and processes. The hazard screening process evaluates an identified hazards in terms of the effects on people, both on-site and off-site. The process uses bounding analyses with no credit given for mitigation of an accident with the exception of certain containers meeting DOT specifications. The process is restricted to human safety issues only. Environmental effects are addressed by the environmental program. Interfaces with environmental organizations will be established in order to share information.

  18. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    SciTech Connect (OSTI)

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

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  19. CRAD, Preliminary Documented Safety Analysis- July 25, 2014 (IEA CRAD 31-2, REV. 0)

    Broader source: Energy.gov [DOE]

    This Criteria Review and Approach Document (IEA CRAD 31-2, REV. 0) provides objectives, criteria, and approaches for reviewing Nuclear Facility Preliminary Documented Safety Analysis.

  20. Preliminary Safety Analysis Report (PSAR), The NSLS 200 MeV Linear Electron Accelerator

    SciTech Connect (OSTI)

    Blumberg, L.N.; Ackerman, A.I.; Dickinson, T.; Heese, R.N.; Larson, R.A.; Neuls, C.W.; Pjerov, S.; Sheehan, J.F.

    1993-06-15

    The radiological, fire and electrical hazards posed by a 200 MeV electron Linear Accelerator, which the NSLS Department will install and commission within a newly assembled structure, are addressed in this Preliminary Safety Analysis Report. Although it is clear that this accelerator is intended to be the injector for a future experimental facility, we address only the Linac in the present PSAR since neither the final design nor the operating characteristics of the experimental facility are known at the present time. The fire detection and control system to be installed in the building is judged to be completely adequate in terms of the marginal hazard presented - no combustible materials other than the usual cabling associated with such a facility have been identified. Likewise, electrical hazards associated with power supplies for the beam transport magnets and accelerator components such as the accelerator klystrons and electron gun are classified as marginal in terms of potential personnel injury, cost of equipment lost, program downtime and public impact perceptions as defined in the BNL Environmental Safety and Health Manual and the probability of occurrence is deemed to be remote. No unusual features have been identified for the power supplies or electrical distribution system, and normal and customary electrical safety standards as practiced throughout the NSLS complex and the Laboratory are specified in this report. The radiation safety hazards are similarly judged to be marginal in terms of probability of occurrence and potential injury consequences since, for the low intensity operation proposed - a factor of 25 less than the maximum Linac capability specified by the vendor - the average beam power is only 0.4 watts. The shielding specifications given in this report will give adequate protection to both the general public and nonradiation workers in areas adjacent to the building as well as radiation workers within the controlled access building.

  1. Criticality safety analysis on fissile materials in Fukushima reactor cores

    SciTech Connect (OSTI)

    Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong; Hirano, Fumio

    2013-07-01

    The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

  2. LANL12-RS-107J PYTHON radiography analysis tool final report...

    Office of Scientific and Technical Information (OSTI)

    LANL12-RS-107J PYTHON radiography analysis tool final report for FY15 Citation Details In-Document Search Title: LANL12-RS-107J PYTHON radiography analysis tool final report for ...

  3. LANL13-RS-107J PYTHON Radiography Analysis Tool Final Report...

    Office of Scientific and Technical Information (OSTI)

    LANL13-RS-107J PYTHON Radiography Analysis Tool Final Report for FY13 Citation Details In-Document Search Title: LANL13-RS-107J PYTHON Radiography Analysis Tool Final Report for ...

  4. Processing Exemptions to Nuclear Safety Rules and Approval of Alternative Methods for Documented Safety Analysis

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

    STD-1083-2009 June 2009 DOE STANDARD PROCESSING EXEMPTIONS TO NUCLEAR SAFETY RULES AND APPROVAL OF ALTERNATIVE METHODS FOR DOCUMENTED SAFETY ANALYSES U.S. Department of Energy AREA SAFT Washington D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. NOT MEASUREMENT SENSITIVE This document is available on the Department of Energy Technical Standards Program Web Page at http://www.hss.energy.gov/nuclearsafety/techstds DOE-STD-1083-2009 iii FOREWORD 1. This

  5. DNFSB 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.2: Safety Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes

    Office of Environmental Management (EM)

    EH-4.2.1.2-Criteria Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.2: Software Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 November 2003 Software Quality Assurance Criteria for Safety Analysis Codes November 2003 INTENTIONALLY BLANK ii Software Quality Assurance Criteria

  6. Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

    SciTech Connect (OSTI)

    John D. Bess; J. Blair Briggs; David W. Nigg

    2009-11-01

    One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

  7. Physics Analysis of Microwave Imaging Data from DIII-D & KSTAR. Final Technical Report

    SciTech Connect (OSTI)

    Munsat, Tobin

    2015-12-31

    Final Technical Report of the award entitled Physics Analysis of Microwave Imaging Data from DIII-D & KSTAR

  8. Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings

    SciTech Connect (OSTI)

    DiSabatino, A; Biswas, D; DeMicco, M; Fisher, L E; Hafner, R; Haslam, J; Mok, G; Patel, C; Russell, E

    2007-04-12

    This Packaging Review Guide (PRG) provides guidance for Department of Energy (DOE) review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE Order 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his or her review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. This PRG is generally organized at the section level in a format similar to that recommended in Regulatory Guide 7.9 (RG 7.9). One notable exception is the addition of Section 9 (Quality Assurance), which is not included as a separate chapter in RG 7.9. Within each section, this PRG addresses the technical and regulatory bases for the review, the manner in which the review is accomplished, and findings that are generally applicable for a package that meets the approval standards. This Packaging Review Guide (PRG) provides guidance for DOE review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE O 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. The primary objectives of this PRG are to: (1) Summarize the regulatory requirements for package approval; (2) Describe the technical review procedures by which DOE determines that these requirements have been satisfied; (3) Establish and maintain the quality and uniformity of reviews; (4) Define the base from which to evaluate proposed changes in scope

  9. Urban Integrated Industrial Cogeneration Systems Analysis. Phase II final report

    SciTech Connect (OSTI)

    Not Available

    1984-01-01

    Through the Urban Integrated Industrial Cogeneration Systems Analysis (UIICSA), the City of Chicago embarked upon an ambitious effort to identify the measure the overall industrial cogeneration market in the city and to evaluate in detail the most promising market opportunities. This report discusses the background of the work completed during Phase II of the UIICSA and presents the results of economic feasibility studies conducted for three potential cogeneration sites in Chicago. Phase II focused on the feasibility of cogeneration at the three most promising sites: the Stockyards and Calumet industrial areas, and the Ford City commercial/industrial complex. Each feasibility case study considered the energy load requirements of the existing facilities at the site and the potential for attracting and serving new growth in the area. Alternative fuels and technologies, and ownership and financing options were also incorporated into the case studies. Finally, site specific considerations such as development incentives, zoning and building code restrictions and environmental requirements were investigated.

  10. A risk-informed approach to safety margins analysis

    SciTech Connect (OSTI)

    Curtis Smith; Diego Mandelli

    2013-07-01

    The Risk Informed Safety Margins Characterization (RISMC) Pathway is a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. The model has been tested on the Advanced Test Reactor (ATR) at Idaho National Lab.

  11. Analysis of key safety metrics of thorium utilization in LWRs

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

    Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; Powers, Jeffrey

    2016-04-08

    Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel (233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on conceptsmore » that mix thorium with uranium (UO2 + ThO2) or that add fertile thorium (ThO2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO2 + ThO2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO2 + UO2 and ThO2 + PuO2 against those of UO2 and typical UO2 + PuO2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of

  12. The Independent Technical Analysis Process Final Report 2006-2007.

    SciTech Connect (OSTI)

    Duberstein, Corey; Ham, Kenneth; Dauble, Dennis; Johnson, Gary

    2007-03-01

    The Bonneville Power Administration (BPA) contracted with the Pacific Northwest National Laboratory (PNNL) to provide technical analytical support for system-wide fish passage information (BPA Project No. 2006-010-00). The goal of this project was to produce rigorous technical analysis products using independent analysts and anonymous peer reviewers. This project provided an independent technical source for non-routine fish passage analyses while allowing routine support functions to be performed by other well-qualified entities. The Independent Technical Analysis Process (ITAP) was created to provide non-routine analysis for fish and wildlife agencies and tribes in particular and the public in general on matters related to juvenile and adult salmon and steelhead passage through the mainstem hydrosystem. The process was designed to maintain the independence of analysts and reviewers from parties requesting analyses, to avoid potential bias in technical products. The objectives identified for this project were to administer a rigorous, transparent process to deliver unbiased technical assistance necessary to coordinate recommendations for storage reservoir and river operations that avoid potential conflicts between anadromous and resident fish. Seven work elements, designated by numbered categories in the Pisces project tracking system, were created to define and accomplish project goals as follows: (1) 118 Coordination - Coordinate technical analysis and review process: (a) Retain expertise for analyst/reviewer roles. (b) Draft research directives. (c) Send directive to the analyst. (d) Coordinate two independent reviews of the draft report. (e) Ensure reviewer comments are addressed within the final report. (2) 162 Analyze/Interpret Data - Implement the independent aspects of the project. (3) 122 Provide Technical Review - Implement the review process for the analysts. (4) 132 Produce Annual Report - FY06 annual progress report with Pisces Disseminate (5) 161

  13. 60-Day waste compatibility safety issues and final results for AY-102 grab samples

    SciTech Connect (OSTI)

    Nuzum, J.L.

    1997-01-31

    Four grab samples (2AY-96-15, 2AY-96-16, 2AY-96-17, and 2AY-96-18) were taken from Riser 15D of Tank 241-AY-102 on October 8, 1996, and received by 222-S Laboratory on October 8, 1996. These samples were analyzed in accordance with Compatibility Grab Sampling and Analysis Plan (TSAP) and Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) in support of the Waste Compatibility Program. No notifications were required based on sample results.

  14. Idaho National Engineering Laboratory (INEL) Environmental Restoration (ER) Program Baseline Safety Analysis File (BSAF)

    SciTech Connect (OSTI)

    1995-09-01

    The Baseline Safety Analysis File (BSAF) is a facility safety reference document for the Idaho National Engineering Laboratory (INEL) environmental restoration activities. The BSAF contains information and guidance for safety analysis documentation required by the U.S. Department of Energy (DOE) for environmental restoration (ER) activities, including: Characterization of potentially contaminated sites. Remedial investigations to identify and remedial actions to clean up existing and potential releases from inactive waste sites Decontamination and dismantlement of surplus facilities. The information is INEL-specific and is in the format required by DOE-EM-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports. An author of safety analysis documentation need only write information concerning that activity and refer to BSAF for further information or copy applicable chapters and sections. The information and guidance provided are suitable for: {sm_bullet} Nuclear facilities (DOE Order 5480-23, Nuclear Safety Analysis Reports) with hazards that meet the Category 3 threshold (DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports) {sm_bullet} Radiological facilities (DOE-EM-STD-5502-94, Hazard Baseline Documentation) Nonnuclear facilities (DOE-EM-STD-5502-94) that are classified as {open_quotes}low{close_quotes} hazard facilities (DOE Order 5481.1B, Safety Analysis and Review System). Additionally, the BSAF could be used as an information source for Health and Safety Plans and for Safety Analysis Reports (SARs) for nuclear facilities with hazards equal to or greater than the Category 2 thresholds, or for nonnuclear facilities with {open_quotes}moderate{close_quotes} or {open_quotes}high{close_quotes} hazard classifications.

  15. Tank Farms Technical Safety Requirements [VOL 1 and 2

    SciTech Connect (OSTI)

    CASH, R.J.

    2000-12-28

    The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR).

  16. New Methods and Tools to Perform Safety Analysis within RISMC

    SciTech Connect (OSTI)

    Diego Mandelli; Curtis Smith; Cristian Rabiti; Andrea Alfonsi; Robert Kinoshita; Joshua Cogliati

    2013-11-01

    The Risk Informed Safety Margins Characterization (RISMC) Pathway uses a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. What differentiates the RISMC approach from traditional probabilistic risk assessment (PRA) is the concept of safety margin. In PRA, a safety metric such as core damage frequency (CDF) is generally estimated using static fault-tree and event-tree models. However, it is not possible to estimate how close we are to physical safety limits (say peak clad temperature) for most accident sequences described in the PRA. In the RISMC approach, what we want to understand is not just the frequency of an event like core damage, but how close we are (or not) to this event and how we might increase our safety margin through margin management strategies in a Dynamic PRA (DPRA) fashion. This paper gives an overview of methods that are currently under development at the Idaho National Laboratory (INL) with the scope of advance the current state of the art of dynamic PRA.

  17. 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.

  18. 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.

  19. 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.

  20. Static-stress analysis of dual-axis safety vessel

    SciTech Connect (OSTI)

    Bultman, D.H.

    1992-11-01

    An 8-ft-diameter safety vessel, made of HSLA-100 steel, is evaluated to determine its ability to contain the quasi-static residual pressure from a high-explosive (HE) blast. The safety vessel is designed for use with the Dual-Axis Radiographic Hydrotest (DARHT) facility being developed at Los Alamos National Laboratory. A smaller confinement vessel fits inside the safety vessel and contains the actual explosion, and the safety vessel functions as a second layer of containment in the unlikely case of a confinement vessel leak. The safety vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, and the Welding Research Council Bulletin, WRC107. Combined stresses that result from internal pressure and external loads on nozzles are calculated and compared to the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzle components are adequately designed for a static pressure of 830 psi, plus the maximum expected external loads. Shell stresses at the shellto-nozzle interface, produced from external loads on the nozzles, were less than 700 psi. The maximum combined stress resulting from the internal pressure plus external loads was 17,384 psi, which is significantly less than the allowable stress of 42,375 psi for HSLA-100 steel.

  1. DOE Standard 3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis, Roll Out Training

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Safety is performing a series of site visits to provide roll-out training and assistance to Program and Site Offices and their contractors on effective implementation of the new revision to DOE Standard 3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis.

  2. Experience in the use of probabilistic safety analysis for the development of safety concepts for commercial high-temperature reactors

    SciTech Connect (OSTI)

    Kroger, W.; Wolters, J.P.

    1986-07-01

    Advanced nuclear reactors in the Federal Republic of Germany (FRG) must analogously fulfill the deterministic safety criteria developed for the light water reactor (LWR). In earlier high-temperature reactor (HTR) concepts, the interpretation of this requirement led to exaggerated safety precautions. Efforts are being made in recent HTR concepts to develop a more specific safety concept making use of probabilistic risk assessment and probabilistic safety analysis. The basis for development and evaluation is formed by a requirement concept of frequency-oriented limits. Design-relevant accidents are divided into three categories and the appertaining maximum permissible doses are allocated based on the FRG Radiation Protection Ordinance. For even more infrequent events it must be demonstrated that the collective damage and risks remain clearly below those of a comparably large LWR. This probabilistic requirement concept has been applied to two HTR concepts under development and the result has been judged positively by a group of experts established by the Federal Ministry of the Interior. The most extensive experience for HTR-500 is discussed in detail.

  3. Final safety evaluation report related to the certification of the System 80+ design: Docket Number 52-002. Supplement 1

    SciTech Connect (OSTI)

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the System 80+ standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1462 in August 1994 to document the NRC staff`s review of the System 80+ design. The System 80+ design was submitted by Asea Brown Boveri-Combustion Engineering (ABB-CE), in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the System 80+ design documentation since the issuance of the FSER. ABB-CE made these changes as a result of its review of the System 80+ design details. The NRC staff concludes that the changes to the System 80+ design documentation are acceptable, and that ABB-CE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the System 80+ design.

  4. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect (OSTI)

    1997-05-01

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  5. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    SciTech Connect (OSTI)

    Herborn, D.I.

    1991-10-01

    The requirements for Westinghouse Hanford independent review of the Preliminary Safety Analysis Report (PSAR) are contained in Section 1.0, Subsection 4.3 of WCH-CM-4-46. Specifically, this manual requires the following: (1) Formal functional reviews of the HWVP PSAR by the future operating organization (HWVP Operations), and the independent review organizations (HWVP and Environmental Safety Assurance, Environmental Assurance, and Quality Assurance); and (2) Review and approval of the HWVP PSAR by the Tank Waste Disposal (TWD) Subcouncil of the Safety and Environmental Advisory Council (SEAC), which provides independent advice to the Westinghouse Hanford President and executives on matters of safety and environmental protection. 7 refs.

  6. Overview of the preliminary safety analysis of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.; McLouth, L.; Odell, B.

    1997-06-01

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF. 10 refs., 6 figs., 4 tabs.

  7. Safety of natural gas dual-fueled vehicles: Addendum to safety analysis of natural gas vehicles transiting highway tunnels

    SciTech Connect (OSTI)

    Shaaban, S.H.; Zalak, V.M. )

    1991-01-01

    A safety analysis was performed to assess the relative hazard of vehicles containing both compressed natural gas (CNG) and gasoline, referred to as dual-fueled vehicles, compared to the hazard of a dedicated CNG vehicle. This study expands upon previous work that examined the safety of CNG vehicles transiting highway tunnels. The approach was to examine operational data, test results and to perform thermal analyses to determine if there are any synergistic effects where the total consequences of fuel release might be greater than the sum of the two fuels released separately. This study concluded that a dual-fueled vehicle poses a slightly greater risk than a dedicated CNG vehicle; however, this marginal increase in risk is small and is within the bounds of risk posed by gasoline-powered vehicles. 4 refs.

  8. 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

  9. Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports

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

    1997-12-12

    he purpose of this DOE Standard is to establish guidance for the preparation and review of hazard categorization and accident analyses techniques as required in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  10. Enterprise Assessments Review of the Los Alamos National Laboratory Transuranic Waste Facility 90% Draft Documented Safety Analysis and Technical Safety Requirements Submittals – EA-LANL-2015-07-07

    Office of Energy Efficiency and Renewable Energy (EERE)

    Operational Awareness Record for the Review of the Transuranic Waste Facility 90% Draft Documented Safety Analysis and Technical Safety Requirements Submittals

  11. ANALYSIS OF THE TANK 6F FINAL CHARACTERIZATION SAMPLES-2012

    SciTech Connect (OSTI)

    Oji, L.; Diprete, D.; Coleman, C.; Hay, M.; Shine, G.

    2012-06-28

    The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm-243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite

  12. Analysis of the Tank 6F Final Characterization Samples-2012

    SciTech Connect (OSTI)

    Oji, L. N.; Diprete, D. P.; Coleman, C. J.; Hay, M. S.; Shine, E. P.

    2013-01-31

    The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm- 243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite

  13. Analysis Of The Tank 6F Final Characterization Samples-2012

    SciTech Connect (OSTI)

    Oji, L. N.; Diprete, D. P.; Coleman, C. J.; Hay, M. S.; Shine, E. P.

    2012-09-27

    The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm-243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite

  14. ANALYSIS OF THE TANK 5F FINAL CHARACTERIZATION SAMPLES-2011

    SciTech Connect (OSTI)

    Oji, L.; Diprete, D.; Coleman, C.; Hay, M.

    2012-08-03

    The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the

  15. ANALYSIS OF THE TANK 5F FINAL CHARATERIZATION SAMPLES-2011

    SciTech Connect (OSTI)

    Oji, L.; Diprete, D.; Coleman, C.; Hay, M.

    2012-01-20

    The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the

  16. Analysis Of The Tank 5F Final Characterization Samples-2011

    SciTech Connect (OSTI)

    Oji, L. N.; Diprete, D.; Coleman, C. J.; Hay, M. S.

    2012-09-27

    The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the

  17. Oak Ridge National Laboratory site data for safety-analysis report

    SciTech Connect (OSTI)

    Fitzpatrick, F.C.

    1982-12-01

    The Oak Ridge National Laboratory site data contained herein were compiled in support of the United States Department of Energy (USDOE) Oak Ridge Operations Office Order OR 5481.1. That order sets forth assignment of responsibilities for safety analysis and review responsibilities and provides guidance relative to the content and format of safety analysis reports. The information presented in this document is intended for use by reference in individual safety analysis reports where applicable to support accident analyses or the establishment of design bases of significance to safety, and it is applicable only to Oak Ridge National Laboratory facilities in Bethel and Melton Valleys. This information includes broad descriptions of the site characteristics, radioactive waste handling and monitoring practices, and the organization and operating policies at Oak Ridge National Laboratory. The historical background of the Laboratory is discussed briefly and the overall physical situation of the facilities is described in the following paragraphs.

  18. Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

    SciTech Connect (OSTI)

    Smith, Curtis L; Mandelli, Diego; Zhegang Ma

    2014-11-01

    As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

  19. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    SciTech Connect (OSTI)

    T. J. Rodovsky

    2010-10-25

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3.

  20. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report. Revision 2

    SciTech Connect (OSTI)

    Herborn, D.I.

    1992-10-01

    This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals for the Hanford Waste Vitrification Plant.

  1. Methodology assessment and recommendations for the Mars science laboratory launch safety analysis.

    SciTech Connect (OSTI)

    Sturgis, Beverly Rainwater; Metzinger, Kurt Evan; Powers, Dana Auburn; Atcitty, Christopher B.; Robinson, David B; Hewson, John C.; Bixler, Nathan E.; Dodson, Brian W.; Potter, Donald L.; Kelly, John E.; MacLean, Heather J.; Bergeron, Kenneth Donald; Bessette, Gregory Carl; Lipinski, Ronald J.

    2006-09-01

    thousands of possible event sequences and to build up a statistical representation of the releases for each accident case. A code to carry out this process will have to be developed or adapted from previous MMRTG missions. Finally, Level C translates the release (or ''source term'') information from Level B into public risk by applying models for atmospheric transport and the health consequences of exposure to the released plutonium dioxide. A number of candidate codes for this level of analysis are available. This report surveys the range of available codes and tools for each of these levels and makes recommendations for which choices are best for the MSL mission. It also identities areas where improvements to the codes are needed. In some cases a second tier of codes may be identified to provide supporting or clarifying insight about particular issues. The main focus of the methodology assessment is to identify a suite of computational tools that can produce a high quality SAR that can be successfully reviewed by external bodies (such as the Interagency Nuclear Safety Review Panel) on the schedule established by NASA and DOE.

  2. FAQS Gap Analysis Qualification Card – Nuclear Explosive Safety Study

    Broader source: Energy.gov [DOE]

    Functional Area Qualification Standard Gap Analysis Qualification Cards outline the differences between the last and latest version of the FAQ Standard.

  3. FAQS Gap Analysis Qualification Card – Senior Technical Safety Manager

    Broader source: Energy.gov [DOE]

    Functional Area Qualification Standard Gap Analysis Qualification Cards outline the differences between the last and latest version of the FAQ Standard.

  4. FAQS Gap Analysis Qualification Card – Criticality Safety

    Broader source: Energy.gov [DOE]

    Functional Area Qualification Standard Gap Analysis Qualification Cards outline the differences between the last and latest version of the FAQ Standard.

  5. Regulatory impact analysis for the petroleum refineries NESHAP. Final report

    SciTech Connect (OSTI)

    1995-08-01

    A regulatory impact analysis (RIA) of the industries affected by the Petroleum Refineries National Emissions Standard for Hazardous Air Pollutants (NESHAP) was completed in support of this regulation. This (RIA) was required because the proposal is economically significant according to Executive Order 12866. The industry for which these impacts was computed was the petroleum refinery industry. Several different impact analyses were included in total or summarized in different chapters in the document. Those analyses were: the compliance cost analysis, the economic impact analysis, and the benefits analysis. Benefits and costs were then compared and discussed in the document`s last chapter.

  6. RISMC advanced safety analysis working plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  7. Risk-Informed Safety Margin Characterization (RISMC): Integrated Treatment of Aleatory and Epistemic Uncertainty in Safety Analysis

    SciTech Connect (OSTI)

    R. W. Youngblood

    2010-10-01

    The concept of “margin” has a long history in nuclear licensing and in the codification of good engineering practices. However, some traditional applications of “margin” have been carried out for surrogate scenarios (such as design basis scenarios), without regard to the actual frequencies of those scenarios, and have been carried out with in a systematically conservative fashion. This means that the effectiveness of the application of the margin concept is determined in part by the original choice of surrogates, and is limited in any case by the degree of conservatism imposed on the evaluation. In the RISMC project, which is part of the Department of Energy’s “Light Water Reactor Sustainability Program” (LWRSP), we are developing a risk-informed characterization of safety margin. Beginning with the traditional discussion of “margin” in terms of a “load” (a physical challenge to system or component function) and a “capacity” (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system response. For example, the probabilistic load spectrum will reflect the frequency of challenges of a particular severity. Such a characterization is required if decision-making is to be informed optimally. However, in order to enable the quantification of probabilistic load spectra, existing analysis capability needs to be extended. Accordingly, the INL is working on a next-generation safety analysis capability whose design will allow for much more efficient parameter uncertainty analysis, and will enable a much better integration of reliability-related and phenomenology-related aspects of margin.

  8. Office of Environmental Protection, Sustainability Support, and Corporate Safety Analysis

    Broader source: Energy.gov [DOE]

    The Office of Environmental Protection, Sustainability Support and Analysis establishes environmental protection requirements and expectations for the Department to ensure protection of workers and the public and protection of the environment from the hazards associated with all Department operations.

  9. ARIES-ACT1 Safety Design and Analysis

    SciTech Connect (OSTI)

    Humrickhouse, Paul W.; Merrill, Brad J.

    2014-01-01

    ARIES-ACT1 (Advanced and Conservative Tokamak) is a 1000-MW(electric) tokamak design featuring advanced plasma physics and divertor and blanket engineering. Some relevant features include an advanced SiC blanket with PbLi 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. We consider here some safety aspects of the ARIES-ACT1 design and model a series of design-basis and beyond-design-basis accidents with the MELCOR code modified for fusion. The presence of multiple coolants (PbLi, helium, and water) makes possible a variety of such accidents. We consider here a loss-of-flow accident caused by a long-term station blackout (LTSBO), an ex-vessel helium break into the cryostat, and a beyond-design-basis accident in which a LTSBO is aggravated by a loss-of-coolant accident in ARIES-ACT1's ultimate decay heat removal system, the water-cooled shield. In the design-basis accidents, we find that the secondary confinement boundaries are not challenged, and the structural integrity of in-vessel components is not threatened by high temperatures or pressures; decay heat can be passively removed.

  10. An advanced deterministic method for spent fuel criticality safety analysis

    SciTech Connect (OSTI)

    DeHart, M.D.

    1998-01-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, non-orthogonal configurations or fissile materials, typical of real world problems. Over the last few years, however, interest in determinist transport methods has been revived, due shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constraints in finite differencing schemes have made discrete ordinates methods impractical for non-orthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitations of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built upon the ESC formalism, is being developed as part of the SCALE code system. This paper will demonstrate the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

  11. Stage Right operational safety analysis and evaluation of Pantex personnel operations

    SciTech Connect (OSTI)

    Rountree, S.L.K.; Whitehurst, H.O.; Tomlin, E.H.; Restrepo, L.F.; White, J. |

    1995-01-01

    This report documents a study (Stage Right Operational Safety Analysis) that was performed to evaluate the effects of new Stage Right operations on the safety of Pantex personnel who perform the operations and maintain the equipment. The primary concern of the evaluation was for personnel safety during Stage Right operations, but operations equipment damage and degradation also were taken into account. This analysis evaluates safety of the work process in the staging of dismantled nuclear weapon pits within the modified Richmond magazines only. This Stage Right Process and Operational Safety Analysis includes the following processes: moving the pelletized drums from the pallet trailer to the pallet turner, staging of pallets and removal of pallets from the magazine, recovery from an incident in a magazine, setting up, opening, and closing a Zone 4 magazine, inventory of pelletized drums in the magazines, transporting pelletized drums from Zone 12 to Zone 4, and maintenance on the shielded lift truck that involves removal of the cab shielding. The analysis includes the following undesirable consequences: injury to personnel, breach of an AL-R8 container, drop of a loaded pallet, damage to equipment, and equipment unreliability.

  12. Station Blackout: A case study in the interaction of mechanistic and probabilistic safety analysis

    SciTech Connect (OSTI)

    Curtis Smith; Diego Mandelli; Cristian Rabiti

    2013-11-01

    The ability to better characterize and quantify safety margins is important to improved decision making about nuclear power plant design, operation, and plant life extension. As research and development (R&D) in the light-water reactor (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. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario.

  13. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    SciTech Connect (OSTI)

    1994-10-01

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results.

  14. LESSONS LEARNED IN DEVELOPMENT OF THE HANFORD SWOC MASTER DOCUMENTED SAFETY ANALYSIS (MDSA) & IMPLEMENTATION VALIDATION REVIEW (IVR)

    SciTech Connect (OSTI)

    MORENO, M.R.

    2004-04-02

    DOE set clear expectations on a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (20 CFR 830, Nuclear Safety Rule), which ensured long-term benefit to Hanford, via issuance of a nuclear safety strategy in February 2003. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development with the goal of a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was approved to standardize methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was approved for the evaluation of radiological consequences for accident scenarios often postulated at Hanford. Standard safety management program chapters were approved for use as a means of compliance with the programmatic chapters of DOE-STD-3009, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports''. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. The new Documented Safety Analysis (DSA) developed to address the operations of four facilities within the Solid Waste Operations Complex (SWOC) necessitated development of an Implementation Validation Review (IVR) process. The IVR process encompasses the following objectives: safety basis controls and requirements are adequately incorporated into appropriate facility documents and work instructions, facility personnel are knowledgeable of controls and requirements, and the DSA/TSR controls have been implemented. Based on DOE direction and safety analysis tools, four waste management nuclear facilities were integrated into one safety basis document. With successful completion of implementation of this safety document, lessons-learned from the in-process review, safety analysis tools and IVR process were documented for future action

  15. Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1

    SciTech Connect (OSTI)

    Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S.

    1997-10-01

    Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.

  16. YUCCA MOUNTAIN SITE CHARACTERIZATION PROJECT EAST-WEST DRIFT SYSTEM SAFETY ANALYSIS

    SciTech Connect (OSTI)

    NA

    1999-06-08

    The purpose of this analysis is to systematically identify and evaluate hazards related to the design of the Yucca Mountain Project Exploratory Studies Facility (ESF) East-West Cross Drift. This analysis builds upon prior ESF System Safety Analyses and incorporates TS Main Drift scenarios, where applicable, into the East-West Drift scenarios. This System Safety Analysis (SSA) focuses on the personnel safety and health hazards associated with the engineered design of the East-West Drift. The analysis also evaluates other aspects of the East-West Drift, including purchased equipment (e.g., scientific mapping platform) or Systems/Structures/Components (SSCs) and out-of-tolerance conditions. In addition to recommending design mitigation features, the analysis identifies the potential need for procedures, training, or Job Safety Analyses (JSAs). The inclusion of this information in the SSA is intended to assist the organization(s) (e.g., constructor, Safety and Health, design) responsible for these aspects of the East-West Drift in evaluating personnel hazards and augment the information developed by these organizations. The SSA is an integral part of the systems engineering process, whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach is used which incorporates operating experiences and recommendations from vendors, the constructor and the operating contractor. The risk assessment in this analysis characterizes the scenarios associated with East-West Drift SSCs in terms of relative risk and includes recommendations for mitigating all identified hazards. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into SSC designs. (2) Add safety features and capabilities to existing designs. (3) Develop procedures and conduct training to increase worker awareness of potential hazards, reduce exposure to hazards, and inform personnel of the

  17. Hybrid methods for cybersecurity analysis : LDRD final report.

    SciTech Connect (OSTI)

    Davis, Warren Leon,; Dunlavy, Daniel M.

    2014-01-01

    Early 2010 saw a signi cant change in adversarial techniques aimed at network intrusion: a shift from malware delivered via email attachments toward the use of hidden, embedded hyperlinks to initiate sequences of downloads and interactions with web sites and network servers containing malicious software. Enterprise security groups were well poised and experienced in defending the former attacks, but the new types of attacks were larger in number, more challenging to detect, dynamic in nature, and required the development of new technologies and analytic capabilities. The Hybrid LDRD project was aimed at delivering new capabilities in large-scale data modeling and analysis to enterprise security operators and analysts and understanding the challenges of detection and prevention of emerging cybersecurity threats. Leveraging previous LDRD research e orts and capabilities in large-scale relational data analysis, large-scale discrete data analysis and visualization, and streaming data analysis, new modeling and analysis capabilities were quickly brought to bear on the problems in email phishing and spear phishing attacks in the Sandia enterprise security operational groups at the onset of the Hybrid project. As part of this project, a software development and deployment framework was created within the security analyst work ow tool sets to facilitate the delivery and testing of new capabilities as they became available, and machine learning algorithms were developed to address the challenge of dynamic threats. Furthermore, researchers from the Hybrid project were embedded in the security analyst groups for almost a full year, engaged in daily operational activities and routines, creating an atmosphere of trust and collaboration between the researchers and security personnel. The Hybrid project has altered the way that research ideas can be incorporated into the production environments of Sandias enterprise security groups, reducing time to deployment from months and

  18. Economic impact analysis for the petroleum refineries NESHAP. Final report

    SciTech Connect (OSTI)

    1995-08-01

    An economic analysis of the industries affected by the Petroleum Refineries National Emmissions Standard for Hazardous Air Pollutants (NESHAP) was completed in support of this standard. The industry for which economic impacts was computed was the petroleum refinery industry. Affected refineries must reduce HAP emissions by the level of control required in the standard. Several types of economic impacts, among them price product changes, output changes, job impacts, and effects on foriegn trade, were computed for the selected regulatory alternative.

  19. 2007 Wholesale Power Rate Case Final Proposal : Risk Analysis Study.

    SciTech Connect (OSTI)

    United States. Bonneville Power Administration.

    2006-07-01

    BPA's operating environment is filled with numerous uncertainties, and thus the rate-setting process must take into account a wide spectrum of risks. The objective of the Risk Analysis is to identify, model, and analyze the impacts that key risks have on BPA's net revenue (total revenues less total expenses). This is carried out in two distinct steps: a risk analysis step, in which the distributions, or profiles, of operating and non operating risks are defined, and a risk mitigation step, in which different rate tools are tested to assess their ability to recover BPA's costs in the face of this uncertainty. Two statistical models are used in the risk analysis step for this rate proposal, the Risk Analysis Model (RiskMod), and the Non-Operating Risk Model (NORM), while a third model, the ToolKit, is used to test the effectiveness of rate tools options in the risk mitigation step. RiskMod is discussed in Sections 2.1 through 2.4, the NORM is discussed in Section 2.5, and the ToolKit is discussed in Section 3. The models function together so that BPA can develop rates that cover all of its costs and provide a high probability of making its Treasury payments on time and in full during the rate period. By law, BPA's payments to Treasury are the lowest priority for revenue application, meaning that payments to Treasury are the first to be missed if financial reserves are insufficient to pay all bills on time. For this reason, BPA measures its potential for recovering costs in terms of probability of being able to make Treasury payments on time (also known as Treasury Payment Probability or TPP).

  20. Simulation for analysis and control of superplastic forming. Final report

    SciTech Connect (OSTI)

    Zacharia, T.; Aramayo, G.A.; Simunovic, S.; Ludtka, G.M.; Khaleel, M.; Johnson, K.I.; Smith, M.T.; Van Arsdale, G.L.; Lavender, C.A.

    1996-08-01

    A joint study was conducted by Oak Ridge National Laboratory (ORNL) and the Pacific Northwest Laboratory (PNL) for the U.S. Department of Energy-Lightweight Materials (DOE-LWM) Program. the purpose of the study was to assess and benchmark the current modeling capabilities with respect to accuracy of predictions and simulation time. Two modeling capabilities with respect to accuracy of predictions and simulation time. Two simulation platforms were considered in this study, which included the LS-DYNA3D code installed on ORNL`s high- performance computers and the finite element code MARC used at PNL. both ORNL and PNL performed superplastic forming (SPF) analysis on a standard butter-tray geometry, which was defined by PNL, to better understand the capabilities of the respective models. The specific geometry was selected and formed at PNL, and the experimental results, such as forming time and thickness at specific locations, were provided for comparisons with numerical predictions. Furthermore, comparisons between the ORNL simulation results, using elasto-plastic analysis, and PNL`s results, using rigid-plastic flow analysis, were performed.

  1. K Basin sludge packaging design criteria (PDC) and safety analysis report for packaging (SARP) approval plan

    SciTech Connect (OSTI)

    Brisbin, S.A.

    1996-03-06

    This document delineates the plan for preparation, review, and approval of the Packaging Design Crieteria for the K Basin Sludge Transportation System and the Associated on-site Safety Analysis Report for Packaging. The transportation system addressed in the subject documents will be used to transport sludge from the K Basins using bulk packaging.

  2. Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report

    SciTech Connect (OSTI)

    Not Available

    1988-06-01

    ''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the fequency of loss of offsite power; the probability that emergency or onsite AC power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of AC power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without AC power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events.

  3. Advances in coupled safety modeling using systems analysis and high-fidelity methods.

    SciTech Connect (OSTI)

    Fanning, T. H.; Thomas, J. W.; Nuclear Engineering Division

    2010-05-31

    The potential for a sodium-cooled fast reactor to survive severe accident initiators with no damage has been demonstrated through whole-plant testing in EBR-II and FFTF. Analysis of the observed natural protective mechanisms suggests that they would be characteristic of a broad range of sodium-cooled fast reactors utilizing metal fuel. However, in order to demonstrate the degree to which new, advanced sodium-cooled fast reactor designs will possess these desired safety features, accurate, high-fidelity, whole-plant dynamics safety simulations will be required. One of the objectives of the advanced safety-modeling component of the Reactor IPSC is to develop a science-based advanced safety simulation capability by utilizing existing safety simulation tools coupled with emerging high-fidelity modeling capabilities in a multi-resolution approach. As part of this integration, an existing whole-plant systems analysis code has been coupled with a high-fidelity computational fluid dynamics code to assess the impact of high-fidelity simulations on safety-related performance. With the coupled capabilities, it is possible to identify critical safety-related phenomenon in advanced reactor designs that cannot be resolved with existing tools. In this report, the impact of coupling is demonstrated by evaluating the conditions of outlet plenum thermal stratification during a protected loss of flow transient. Outlet plenum stratification was anticipated to alter core temperatures and flows predicted during natural circulation conditions. This effect was observed during the simulations. What was not anticipated, however, is the far-reaching impact that resolving thermal stratification has on the whole plant. The high temperatures predicted at the IHX inlet due to thermal stratification in the outlet plenum forces heat into the intermediate system to the point that it eventually becomes a source of heat for the primary system. The results also suggest that flow stagnation in the

  4. Final Report: Weatherization and Energy Conservation Education and Home Energy and Safety Review in the Aleutian Islands

    SciTech Connect (OSTI)

    Bruce Wright

    2011-08-30

    Aleutian/Pribilof Islands Association, Inc. (APIA) hired three part-time local community members that desire to be Energy Technicians. The energy technicians were trained in methods of weatherization assistance, energy conservation and home safety. They developed a listing of homes in the region that required weatherization, and conducted on-site weatherization and energy conservation education and a home energy and safety reviews in the communities of Akutan, False Pass, King Cove and Nelson Lagoon. Priority was given to these smaller communities as they tend to have the residences most in need of weatherization and energy conservation measures. Local residents were trained to provide all three aspects of the project: weatherization, energy conservation education and a home energy and safety review. If the total energy saved by installing these products is a 25% reduction (electrical and heating, both of which are usually produced by combustion of diesel fuel), and the average Alaska home produces 32,000 pounds of CO2 each year, so we have saved about: 66 homes x 16 tons of CO2 each year x .25 = 264 tons of CO2 each year.

  5. Facility Safety

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

    2012-12-04

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. This Page Change is limited in scope to changes necessary to invoke DOE-STD-1104, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Document, and revised DOE-STD-3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis as required methods. DOE O 420.1C Chg 1, dated 2-27-15, supersedes DOE O 420.1C.

  6. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    SciTech Connect (OSTI)

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  7. An object-oriented approach to risk and reliability analysis : methodology and aviation safety applications.

    SciTech Connect (OSTI)

    Dandini, Vincent John; Duran, Felicia Angelica; Wyss, Gregory Dane

    2003-09-01

    This article describes how features of event tree analysis and Monte Carlo-based discrete event simulation can be combined with concepts from object-oriented analysis to develop a new risk assessment methodology, with some of the best features of each. The resultant object-based event scenario tree (OBEST) methodology enables an analyst to rapidly construct realistic models for scenarios for which an a priori discovery of event ordering is either cumbersome or impossible. Each scenario produced by OBEST is automatically associated with a likelihood estimate because probabilistic branching is integral to the object model definition. The OBEST methodology is then applied to an aviation safety problem that considers mechanisms by which an aircraft might become involved in a runway incursion incident. The resulting OBEST model demonstrates how a close link between human reliability analysis and probabilistic risk assessment methods can provide important insights into aviation safety phenomenology.

  8. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect (OSTI)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  9. MODEL 9977 B(M)F-96 SAFETY ANALYSIS REPORT FOR PACKAGING

    SciTech Connect (OSTI)

    Abramczyk, G; Paul Blanton, P; Kurt Eberl, K

    2006-05-18

    This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on and for the 9977 Shipping Package, referred to as the General Purpose Fissile Package (GPFP). The performance evaluation presented in this SARP documents the compliance of the 9977 package with the regulatory safety requirements for Type B packages. Per 10 CFR 71.59, for the 9977 packages evaluated in this SARP, the value of ''N'' is 50, and the Transport Index based on nuclear criticality control is 1.0. The 9977 package is designed with a high degree of single containment. The 9977 complies with 10 CFR 71 (2002), Department of Energy (DOE) Order 460.1B, DOE Order 460.2, and 10 CFR 20 (2003) for As Low As Reasonably Achievable (ALARA) principles. The 9977 also satisfies the requirements of the Regulations for the Safe Transport of Radioactive Material--1996 Edition (Revised)--Requirements. IAEA Safety Standards, Safety Series No. TS-R-1 (ST-1, Rev.), International Atomic Energy Agency, Vienna, Austria (2000). The 9977 package is designed, analyzed and fabricated in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1992 edition.

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  11. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  12. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    SciTech Connect (OSTI)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

    2014-02-12

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  13. Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets

    SciTech Connect (OSTI)

    Downar, Thomas; Zazimi, Mujid; Hill, Bob

    2015-01-31

    The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.

  14. A probabilistic safety analysis of UF{sub 6} handling at the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Boyd, G.J.; Lewis, S.R.; Summitt, R.L.

    1991-12-31

    A probabilistic safety study of UF{sub 6} handling activities at the Portsmouth Gaseous Diffusion Plant has recently been completed. The analysis provides a unique perspective on the safety of UF{sub 6} handling activities. The estimated release frequencies provide an understanding of current risks, and the examination of individual contributors yields a ranking of important plant features and operations. Aside from the probabilistic results, however, there is an even more important benefit derived from a systematic modeling of all operations. The integrated approach employed in the analysis allows the interrelationships among the equipment and the required operations to be explored in depth. This paper summarizes the methods used in the study and provides an overview of some of the technical insights that were obtained. Specific areas of possible improvement in operations are described.

  15. Seismic safety margins research program. Phase I. Final report: plant/site selection and data collection (Project I)

    SciTech Connect (OSTI)

    Chuang, T. Y.

    1981-05-01

    Project I of Phase I of the Seismic Safety Margins Research Program (SSMRP) comprised two parts: the selection of a representative nuclear power plant/site for study in Phase I and the collection of data needed by the other SSMRP projects. Unit 1 of the Zion Nuclear Power Plant in Zion, Illinois, was selected for the SSMRP Phase I studies. The Zion plant and its site were found to be reasonably representative of operating and future plants with regard to its nuclear steam supply system; the type of containment structure (prestressed concrete); its electrical capacity (1100 MWe); its location (the Midwest); the peak seismic accelaration used for design (0.17g); and the properties of the underlying soil (the low-strain shear-wave velocity is 1650 ft/s in a 50- to 100-ft-thick layer of soil overlying sedimentary bedrock).

  16. Preparation of Documented Safety Analysis for Interim Operations at DOE Nuclear Facilities

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

    3011-2016 January 2016 DOE STANDARD PREPARATION OF DOCUMENTED SAFETY ANALYSIS FOR INTERIM OPERATIONS AT DOE NUCLEAR FACILITIES U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-STD-3011-2016 ii FOREWORD 1. This Department of Energy (DOE) Standard (STD) has been approved to be used by DOE, including the National Nuclear Security Administration, and their contractors. 2. Beneficial comments

  17. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report. Revision 3

    SciTech Connect (OSTI)

    Herborn, D.I.

    1993-11-01

    Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals.

  18. Pedigree Analysis of the MELCOR 1.8.2 Code to be Used for ITER’s Report Preliminary on Safety

    SciTech Connect (OSTI)

    Richard L. Moore; Brad J. Merrill

    2007-06-01

    This report documents the pedigree analysis of the MELCOR 1.8.2 code to be used for ITER’s Report Preliminary on Safety. To pedigree the code the process involved four steps. First, taking the modified MELCOR 1.8.2 code used by the ITER Joint Central Team (JCT) for analyses in previous ITER Safety Assessments and compared the FORTRAN code of this version line-by-line to the original 1.8.2 version of MELCOR. The second step was a non-regression analysis which involves comparing the results from the pedigreed version against those predicted by the original, unmodified version of MELCOR 1.8.2. The third step involved comparing the pedigreed version results to results from the MELCOR version used by the ITER JCT for the Generic Site Safety Report (GSSR) against a set of accident problems analyzed for the safety report. The fourth and final step involved a comparison between the pedigreed version of the code and the developmental test problems cited in the change documents referenced in this report. The results from the pedigree process are described in this report.

  19. An analysis of tank and pump pit flammable gas data in support of saltwater pumping safety basis simplification

    SciTech Connect (OSTI)

    MCCAIN, D.J.

    2000-07-26

    Hanford Site high-level waste tanks are interim stabilized by pumping supernatant and interstitial waste liquids to double-shell tanks (DSTs) through a saltwell pump (SWP). The motor to this SWP is located atop the tank, inside a pump pit. A pumping line extends down from the pump motor into the well area, located in the salt/sludge solids in the tank below. Pumping of these wastes is complicated by the fact that some of the wastes generate and retain potentially hazardous amounts of hydrogen, nitrous oxide, and ammonia. Monitoring of flammable gas concentrations during saltwell pumping activities has shown that one effect of pumping is acceleration in the release of accumulated hydrogen. A second effect is that of a temporarily increased hydrogen concentration in both the dome space and pump pit. There is a safety concern that the hydrogen concentration during saltwell pumping activities might approach the lower flammability limit (LFL) in either the tank dome space or the pump pit. The current Final Safety Analysis Report (FSAR) (CHG 2000) for saltwell pumping requires continuous flammable gas monitoring in both the pump pit and the tank vapor space during saltwell pumping. The FSAR also requires that portable exhauster fans be available by most of the passively ventilated tanks to be saltwell pumped in the event that additional air flow is required to dilute the headspace concentration of flammable gases to acceptable levels. The first objective of this analysis is to review the need for an auxiliary exhauster. Since the purpose of the exhauster is to diffuse unacceptably high flammable gas concentrations, discovery of an alternate method of accomplishing the same task may provide cost savings. The method reviewed is that of temporarily stopping the saltwell pumps. This analysis also examines the typical hydrogen concentration peaks and the rates of increase in hydrogen levels already witnessed in tanks during saltwell pumping activities. The historical data

  20. Enterprise Assessments, Review of the Hanford Site Sludge Treatment Project Engineered Container Retrieval and Transfer System Preliminary Documented Safety Analysis, Revision 00- April 2015

    Office of Energy Efficiency and Renewable Energy (EERE)

    Review of Hanford Site Sludge Treatment Project Engineering Container Retrieval and Transfer System Preliminary Documented Safety Analysis

  1. SAFETY ANALYSIS FOR TANK 241-AZ-101 MIXER PUMP PROCESS TEST

    SciTech Connect (OSTI)

    HAMMOND DM; HARRIS JP; MOUETTE P

    1997-06-09

    This document contains the completed safety analysis which establishes the safety envelope for performing the mixer pump process test in Tank 241-AZ-101. This process test is described in TF-210-OTP-001. All equipment necessary for the mixer pump test has been installed by Project W-151. The purpose of this document is to describe and analyze the mixer pump test for Aging Waste Facility (AWF) Tank 241-AZ-101 and to address the 'yes/maybe' responses marked for evaluation questions identified in Unreviewed Safety Question Evaluation (USQE) TF-94-0266. The scope of this document is limited to the performance of the mixer pump test for Tank 241-AZ-101. Unreviewed Safety Question Determination (USQD) TF-96-0018 verified that the installation of two mixer pumps into Tank 241-AZ-101 was within the current Tank Waste Remediation System (TWRS) Authorization Basis. USQDs TF-96-0461, TF-96-0448, and TF-96-0805 verified that the installation of the in-tank video camera, thermocouples, and Ultrasonic Interface Level Analyzer (URSILLA), respectively, were within the current TWRS Authorization Basis. USQD TF-96-1041 verified that the checkout testing of the installed equipment was within the current TWRS Authorization Basis. Installation of the pumps and equipment has been completed. An evaluation of safety considerations associated with operation of the mixer pumps for the mixer pump test is provided in this document. This document augments the existing AWF authorization basis as defined in the Interim Safety Basis (Stahl 1997), and as such, will use the existing Interim Operational Safety Requirements (IOSRs) of Heubach 1996 to adequately control the mixer pump test. The hazard and accident analysis is limited to the scope and impact of the mixer pump test, and therefore does not address hazards already addressed by the current AWF authorization basis. This document does not evaluate removal of the mixer pumps. Safety considerations for removal of the pumps will be addressed by

  2. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE`s System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E.

  3. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 1, Chapters 1--14

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE`s System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report.

  4. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    SciTech Connect (OSTI)

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  5. 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).

  6. Safety analysis report for packaging a DOT 7A specification container for tritiated liquid wastes

    SciTech Connect (OSTI)

    Alford, E.

    1980-08-01

    This Safety Analysis Report for Packaging (SARP) was prepared in accordance with ERDA (DOE) Appendix 5201 for DOE/ALO review and approval of packaging of tritiated liquid wastes to be shipped from Sandia National Laboratories, Livermore, (SNLL) California. This report presents information pertinent to the construction of tritiated liquid waste shipping containers. It contains design and development considerations, explains tests and evaluations required to prove the container can withstand normal transportation conditions, and demonstrates that the Sandia container-and-radioactive-material shipment package is in compliance with DOE and Department of Transportation (DOT) safety requirements. An internal review of this SARP has been performed in compliance with the ERDA (DOE) Manual, 5201 Appendix V.

  7. Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

    SciTech Connect (OSTI)

    Not Available

    1991-10-18

    The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173.

  8. Frequency agile laser safety & hazard analysis for the Sandia Remote Sensing System LIDAR.

    SciTech Connect (OSTI)

    Augustoni, Arnold L.

    2009-05-01

    A laser safety and hazard analysis was performed for the Raytheon Frequency Agile Laser (FAL) to be used with the Sandia Remote Sensing System (SRSS) B-70 Trailer based on the 2007 version of the American National Standards Institute's (ANSI) Standard 136.1, for Safe Use of Lasers and the 2005 version of the ANSI Standard Z136.6, for Safe Use of Lasers Outdoors. The B-70 SRSS LIDAR system is a portable platform, which is used to perform laser interaction experiments and tests at various national test sites.

  9. Breach and safety analysis of spills over water from large liquefied natural gas carriers.

    SciTech Connect (OSTI)

    Hightower, Marion Michael; Luketa-Hanlin, Anay Josephine; Attaway, Stephen W.

    2008-05-01

    In 2004, at the request of the Department of Energy, Sandia National Laboratories (Sandia) prepared a report, ''Guidance on the Risk and Safety Analysis of Large Liquefied Natural Gas (LNG) Spills Over Water''. That report provided framework for assessing hazards and identifying approaches to minimize the consequences to people and property from an LNG spill over water. The report also presented the general scale of possible hazards from a spill from 125,000 m3 o 150,000 m3 class LNG carriers, at the time the most common LNG carrier capacity.

  10. A Comparison of Modifications to MELCOR versions 1.8.2 and 1.8.6 for ITER Safety Analysis

    SciTech Connect (OSTI)

    B. J. Merrill; P. W. Humrickhouse

    2010-06-01

    -regression analysis that involves comparing the results from the modified version of MELCOR 1.8.6 against those predicted by the original, unmodified version of MELCOR 1.8.6. The purpose of this non-regression analysis is to demonstrate that the modifications made to the MELCOR 1.8.6 code do not drastically alter the intended functions of the MELCOR base code, and if they do to explain why the departure occurs and if the change is needed. The analysis was performed using the demonstration problem that came with the MELCOR 1.8.6 source distribution. Section 3 makes a comparison between the pedigreed version of MELCOR 1.8.2 and the new modified version of MELCOR 1.8.6 on a set of accident problems used by the ITER Joint Central Team (JCT) in the Generic Site Safety Report (GSSR) [9]. Finally, in section 4, the two versions of the code are compared on a series of developmental test problems described in the change documents [2-4, 6].

  11. K Basin Hazard Analysis

    SciTech Connect (OSTI)

    PECH, S.H.

    2000-08-23

    This report describes the methodology used in conducting the K Basins Hazard Analysis, which provides the foundation for the K Basins Final Safety Analysis Report. This hazard analysis was performed in accordance with guidance provided by DOE-STD-3009-94, Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports and implements the requirements of DOE Order 5480.23, Nuclear Safety Analysis Report.

  12. Regulatory/backfit analysis for the resolution of Unresolved Safety Issue A-44, Station Blackout

    SciTech Connect (OSTI)

    Rubin, A.M.

    1988-06-01

    Station blackout is the complete loss of alternating current (ac) electric power to the essential and nonessential buses in a nuclear power plant; it results when both offsite power and the onsite emergency ac power systems are unavailable. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the number of failures of emergency diesel generators, and the potentially severe consequences of a loss of all ac power, ''Station Blackout'' was designated as Unresolved Safety Issue (USI) A-44. This report presents the regulatory/backfit analysis for USI A-44. It includes (1) a summary of the issue, (2) the recommended technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC) staff, (4) an assessment of the benefits and costs of the recommended resolution, (5) the decision rationale, (6) the relationship between USI A-44 and other NRC programs and requirements, and (7) a backfit analysis demonstrating that the resolution of USI A-44 complies with the backfit rule (10 CFR 50.109).

  13. Demonstration of emulator-based Bayesian calibration of safety analysis codes: Theory and formulation

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

    Yurko, Joseph P.; Buongiorno, Jacopo; Youngblood, Robert

    2015-05-28

    System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here withmore » Markov Chain Monte Carlo (MCMC) sampling feasible. This study uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.« less

  14. Demonstration of emulator-based Bayesian calibration of safety analysis codes: Theory and formulation

    SciTech Connect (OSTI)

    Yurko, Joseph P.; Buongiorno, Jacopo; Youngblood, Robert

    2015-05-28

    System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC) sampling feasible. This study uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.

  15. Gradual degradation of concrete fiber containers and preliminary Safety analysis for the Slovak near-surface repository

    SciTech Connect (OSTI)

    Duran, Juraj

    2007-07-01

    Available in abstract form only. Full text of publication follows: National Radioactive Waste Repository will be used for safe disposal of low and intermediate-level radioactive wastes in Mochovce, Slovak Republic. The Preliminary Safety Analysis Report (PSAR) has developed a conceptual model that strongly overestimated radiological exposures for the Normal Evolution Scenario (NES). Radioactive waste management required additional measures for safe disposal of radioactive waste to minimize the potential consequence to the workers and the public. Use of Fiber Reinforced Containers (FRC) is proposed to enhance the performance of the potential repository for safe disposal of radioactive waste and reduce the probability of exposure. This paper contains the description of models, methods, results from experimental measurements and input data, which were used for probabilistic calculations of the lifetime FRC. The Cumulative Distribution Function (CDF) for the FRC lifetime was used to modify the conceptual model for NES. The model assumed gradual degradation of the FRC and gradual intrusion of water to the repository volume. These assumptions are in contrary to the ones in the PSAR that assumed instantaneous degradation of the FRC and instantaneous fill-up of the repository volume. The model showed that the new assumptions resulted in less radiological consequences, which allows for final design modifications of the repository. (author)

  16. Probabilistic risk analysis toward cost-effective 3S (safety, safeguards, security) implementation

    SciTech Connect (OSTI)

    Suzuki, Mitsutoshi; Mochiji, Toshiro

    2014-09-30

    Probabilistic Risk Analysis (PRA) has been introduced for several decades in safety and nuclear advanced countries have already used this methodology in their own regulatory systems. However, PRA has not been developed in safeguards and security so far because of inherent difficulties in intentional and malicious acts. In this paper, probabilistic proliferation and risk analysis based on random process is applied to hypothetical reprocessing process and physical protection system in nuclear reactor with the Markov model that was originally developed by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) in Generation IV International Framework (GIF). Through the challenge to quantify the security risk with a frequency in this model, integrated risk notion among 3S to pursue the cost-effective installation of those countermeasures is discussed in a heroic manner.

  17. Los Alamos National Laboratory corregated metal pipe saw facility preliminary safety analysis report. Volume I

    SciTech Connect (OSTI)

    1990-09-19

    This Preliminary Safety Analysis Report addresses site assessment, facility design and construction, and design operation of the processing systems in the Corrugated Metal Pipe Saw Facility with respect to normal and abnormal conditions. Potential hazards are identified, credible accidents relative to the operation of the facility and the process systems are analyzed, and the consequences of postulated accidents are presented. The risk associated with normal operations, abnormal operations, and natural phenomena are analyzed. The accident analysis presented shows that the impact of the facility will be acceptable for all foreseeable normal and abnormal conditions of operation. Specifically, under normal conditions the facility will have impacts within the limits posted by applicable DOE guidelines, and in accident conditions the facility will similarly meet or exceed the requirements of all applicable standards. 16 figs., 6 tabs.

  18. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  19. ORNL necessary and sufficient standards for environment, safety, and health. Final report of the Identification Team for other industrial, radiological, and non-radiological hazard facilities

    SciTech Connect (OSTI)

    1998-07-01

    This Necessary and Sufficient (N and S) set of standards is for Other Industrial, Radiological, and Non-Radiological Hazard Facilities at Oak Ridge National Laboratory (ORNL). These facility classifications are based on a laboratory-wide approach to classify facilities by hazard category. An analysis of the hazards associated with the facilities at ORNL was conducted in 1993. To identify standards appropriate for these Other Industrial, Radiological, and Non-Radiological Hazard Facilities, the activities conducted in these facilities were assessed, and the hazards associated with the activities were identified. A preliminary hazards list was distributed to all ORNL organizations. The hazards identified in prior hazard analyses are contained in the list, and a category of other was provided in each general hazard area. A workshop to assist organizations in properly completing the list was held. Completed hazard screening lists were compiled for each ORNL division, and a master list was compiled for all Other Industrial, Radiological Hazard, and Non-Radiological facilities and activities. The master list was compared against the results of prior hazard analyses by research and development and environment, safety, and health personnel to ensure completeness. This list, which served as a basis for identifying applicable environment, safety, and health standards, appears in Appendix A.

  20. Probabilistic cost-benefit analysis of enhanced safety features for strategic nuclear weapons at a representative location

    SciTech Connect (OSTI)

    Stephens, D.R.; Hall, C.H.; Holman, G.S.; Graham, K.F.; Harvey, T.F.; Serduke, F.J.D.

    1993-10-01

    We carried out a demonstration analysis of the value of developing and implementing enhanced safety features for nuclear weapons in the US stockpile. We modified an approach that the Nuclear Regulatory Commission (NRC) developed in response to a congressional directive that NRC assess the ``value-impact`` of regulatory actions for commercial nuclear power plants. Because improving weapon safety shares some basic objectives with NRC regulations, i.e., protecting public health and safety from the effects of accidents involving radioactive materials, we believe the NRC approach to be appropriate for evaluating weapons-safety cost-benefit issues. Impact analysis includes not only direct costs associated with retrofitting the weapon system, but also the expected costs (or economic risks) that are avoided by the action, i.e., the benefits.

  1. Analysis and comparison of five contractor safety and health manuals (EG and G, SR II, ORNL, Ashland, and MLGW)

    SciTech Connect (OSTI)

    Crowder, C.; Hurley, T.

    1981-03-01

    An analysis is presented of five safety and health contractor manuals against the requirements of the FE OSH Manual (FE 5480.1), and a breakdown in chart form of how the manuals compare to each other is given. It is pointed out that the manuals are inadequate, but that site visits will be necessary to determine the actual comprehensiveness of the facilities' safety and health programs.

  2. FINAL REPORT

    Office of Scientific and Technical Information (OSTI)

    FINAL REPORT Analytical and Elemental Analysis of Air and Soil Samples Facility and Public ... Information 4 Background 5 Stormwater Pollution 5 Erosion and Sediment Control Workshop ...

  3. Safety analysis of high pressure 3He-filled micro-channels for thermal neutron detection.

    SciTech Connect (OSTI)

    Ferko, Scott M.; Galambos, Paul C.; Derzon, Mark Steven; Renzi, Ronald F.

    2008-11-01

    This document is a safety analysis of a novel neutron detection technology developed by Sandia National Laboratories. This technology is comprised of devices with tiny channels containing high pressure {sup 3}He. These devices are further integrated into large scale neutron sensors. Modeling and preliminary device testing indicates that the time required to detect the presence of special nuclear materials may be reduced under optimal conditions by several orders of magnitude using this approach. Also, these devices make efficient use of our {sup 3}He supply by making individual devices more efficient and/or extending the our limited {sup 3}He supply. The safety of these high pressure devices has been a primary concern. We address these safety concerns for a flat panel configuration intended for thermal neutron detection. Ballistic impact tests using 3 g projectiles were performed on devices made from FR4, Silicon, and Parmax materials. In addition to impact testing, operational limits were determined by pressurizing the devices either to failure or until they unacceptably leaked. We found that (1) sympathetic or parasitic failure does not occur in pressurized FR4 devices (2) the Si devices exhibited benign brittle failure (sympathetic failure under pressure was not tested) and (3) the Parmax devices failed unacceptably. FR4 devices were filled to pressures up to 4000 + 100 psig, and the impacts were captured using a high speed camera. The brittle Si devices shattered, but were completely contained when wrapped in thin tape, while the ductile FR4 devices deformed only. Even at 4000 psi the energy density of the compressed gas appears to be insignificant compared to the impact caused by the incoming projectile. In conclusion, the current FR4 device design pressurized up to 4000 psi does not show evidence of sympathetic failure, and these devices are intrinsically safe.

  4. ANSI/ASHRAE/IESNA Standard 90.1-2007 Final Determination Quantitative Analysis

    SciTech Connect (OSTI)

    Halverson, Mark A.; Liu, Bing; Richman, Eric E.; Winiarski, David W.

    2011-05-01

    The United States (U.S.) Department of Energy (DOE) conducted a final quantitative analysis to assess whether buildings constructed according to the requirements of the American National Standards Institute (ANSI)/American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE)/Illuminating Engineering Society of North America (IESNA) Standard 90.1-2007 would result in energy savings compared with buildings constructed to ANSI/ASHRAE/IESNA Standard 90.1-2004. The final analysis considered each of the 44 addenda to ANSI/ASHRAE/IESNA Standard 90.1-2004 that were included in ANSI/ASHRAE/IESNA Standard 90.1-2007. All 44 addenda processed by ASHRAE in the creation of Standard 90.1-2007 from Standard 90.1-2004 were reviewed by DOE, and their combined impact on a suite of 15 building prototype models in 15 ASHRAE climate zones was considered. Most addenda were deemed to have little quantifiable impact on building efficiency for the purpose of DOE’s final determination. However, out of the 44 addenda, 9 were preliminarily determined to have measureable and quantifiable impact.

  5. Radiological Safety Analysis Computer (RSAC) Program Version 7.2 Users’ Manual

    SciTech Connect (OSTI)

    Dr. Bradley J Schrader

    2010-10-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.2 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users’ manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods.

  6. Radiological Safety Analysis Computer (RSAC) Program Version 7.0 Users’ Manual

    SciTech Connect (OSTI)

    Dr. Bradley J Schrader

    2009-03-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.0 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users’ manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods.

  7. Analysis of human factors effects on the safety of transporting radioactive waste materials: Technical report

    SciTech Connect (OSTI)

    Abkowitz, M.D.; Abkowitz, S.B.; Lepofsky, M.

    1989-04-01

    This report examines the extent of human factors effects on the safety of transporting radioactive waste materials. It is seen principally as a scoping effort, to establish whether there is a need for DOE to undertake a more formal approach to studying human factors in radioactive waste transport, and if so, logical directions for that program to follow. Human factors effects are evaluated on driving and loading/transfer operations only. Particular emphasis is placed on the driving function, examining the relationship between human error and safety as it relates to the impairment of driver performance. Although multi-modal in focus, the widespread availability of data and previous literature on truck operations resulted in a primary study focus on the trucking mode from the standpoint of policy development. In addition to the analysis of human factors accident statistics, the report provides relevant background material on several policies that have been instituted or are under consideration, directed at improving human reliability in the transport sector. On the basis of reported findings, preliminary policy areas are identified. 71 refs., 26 figs., 5 tabs.

  8. The Oak Ridge Research Reactor: Safety analysis: Volume 2, Supplement 3

    SciTech Connect (OSTI)

    Cook, D.H.; Hamrick, T.P.

    1987-06-29

    The Oak Ridge Research Reactor (ORR) was constructed in the mid 1950s. Since it is an older facility, the issue of life-limiting conditions or material deterioration resulting from prolonged exposure to the normal operating environment is an item that should be addressed in the safety analysis for the ORR. Life-limiting conditions were considered in the original design of ORR; but due to the limited data that were available at that time on material performance in research reactors, various studies were completed during the first 10 years of operation at ORR to verify the applicable life-limiting parameters. Based on today's knowledge of life limiting conditions and the previous 30 years of operating experience at the ORR facility, the three specific areas of concern are addressed in this supplement: (1) embrittlement of the structures due to radiation damage, which is described in Section 2; (2) fatigue due to the effects of both thermal cycling and vibration, which is addressed in Section 3; and (3) the effects of corrosion on the integrity of the primary system, which is described in Section 4. The purpose of this document is to provide a review of the applicable safety studies which have been performed, and to state the status of the ORR with regard to embrittlement, fatigue (due to thermal cycling and vibration), and corrosion.

  9. Regulatory analysis for the resolution of Unresolved Safety Issue A-44, Station Blackout. Draft report

    SciTech Connect (OSTI)

    Rubin, A.M.

    1986-01-01

    ''Station Blackout'' is the complete loss of alternating current (ac) electric power to the essential and nonessential buses in a nuclear power plant; it results when both offsite power and the onsite emergency ac power systems are unavailable. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the number of failures of emergency diesel generators, and the potentially severe consequences of a loss of all ac power, ''Station Blackout'' was designated as Unresolved Safety Issue (USI) A-44. This report presents the regulatory analysis for USI A-44. It includes: (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC) staff, (4) an assessment of the benefits and costs of the recommended resolution, (5) the decision rationale, and (6) the relationship between USI A-44 and other NRC programs and requirements.

  10. Safety analysis -- 200 Area Savannah River Plant, F-Canyon Operations. Supplement 4

    SciTech Connect (OSTI)

    Beary, M.M.; Collier, C.D.; Fairobent, L.A.; Graham, R.F.; Mason, C.L.; McDuffee, W.T.; Owen, T.L.; Walker, D.H.

    1986-02-01

    The F-Canyon facility is located in the 200 Separations Area and uses the Purex process to recover plutonium from reactor-irradiated uranium. The irradiated uranium is normally in the form of solid or hollow cylinders called slugs. These slugs are encased in aluminum cladding and are sent to the F-Canyon from the Savannah River Plant (SRP) reactor areas or from the Receiving Basin for Offsite Fuels (RBOF). This Safety Analysis Report (SAR) documents an analysis of the F-Canyon operations and is an update to a section of a previous SAR. The previous SAR documented an analysis of the entire 200 Separations Area operations. This SAR documents an analysis of the F-Canyon and is one of a series of documents for the Separations Area as specified in the Savannah River Implementation Plans. A substantial amount of the information supporting the conclusions of this SAR is found in the Systems Analysis. Some F-Canyon equipment has been updated during the time between the Systems Analysis and this SAR and a complete description of this equipment is included in this report. The primary purpose of the analysis was to demonstrate that the F-Canyon can be operated without undue risk to onsite or offsite populations and to the environment. In this report, risk is defined as the expected frequency of an accident, multiplied by the resulting radiological consequence in person-rem. The units of risk for radiological dose are person-rem/year. Maximum individual exposure values have also been calculated and reported.

  11. 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.

  12. The Radiological Safety Analysis Computer Program (RSAC-5) user`s manual. Revision 1

    SciTech Connect (OSTI)

    Wenzel, D.R.

    1994-02-01

    The Radiological Safety Analysis Computer Program (RSAC-5) calculates the consequences of the release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory from either reactor operating history or nuclear criticalities. RSAC-5 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated through the inhalation, immersion, ground surface, and ingestion pathways. RSAC+, a menu-driven companion program to RSAC-5, assists users in creating and running RSAC-5 input files. This user`s manual contains the mathematical models and operating instructions for RSAC-5 and RSAC+. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-5 and RSAC+. These programs are designed for users who are familiar with radiological dose assessment methods.

  13. Safety analysis report for packaging (onsite) decontaminated equipment self-container

    SciTech Connect (OSTI)

    Boehnke, W.M.

    1998-09-29

    The purpose of this Safety Analysis Report for Packaging (SARP) is to demonstrate that specific decontaminated equipment can be safely used as its own self-container. As a Decontaminated Equipment Self-Container (also referred to as a self-container), no other packaging, such as a burial box, would be required to transport the equipment onsite. The self-container will consist of a piece of equipment or apparatus which has all readily removable interior contamination removed, all of its external openings sealed, and all external surfaces decontaminated to less than 2000 dpm/100 cm for gamma-emitting radionuclides and less than 220 dpm/100 CM2 for alpha-emitting radionuclides.

  14. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    SciTech Connect (OSTI)

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  15. TWRS safety SSCs: Requirements and characteristics

    SciTech Connect (OSTI)

    Smith-Fewell, M.A.

    1997-12-30

    Safety Systems, Structures, and Components (SSCs) have been identified from hazard and accident analyses. These analyses were performed to support the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR) and Basis for Interim Operation (BID). The text identifies and evaluates the SSCs and their supporting SSCs to show that they either prevent the occurrence of the accident or mitigate the consequences of the accident to below the acceptance guidelines. The requirements for the SSCs to fulfill these tasks are described.

  16. Final Report: Interphase Analysis and Control in Fiber Reinforced Thermoplastic Composites

    SciTech Connect (OSTI)

    Jon J. Kellar; William M. Cross; Lidvin Kjerengtroen

    2009-03-14

    This research program builds upon a multi-disciplinary effort in interphase analysis and control in thermoplastic matrix polymer matrix composites (PMC). The research investigates model systems deemed of interest by members of the Automotive Composites Consortium (ACC) as well as samples at the forefront of PMC process development (DRIFT and P4 technologies). Finally, the research investigates, based upon the fundamental understanding of the interphases created during the fabrication of thermoplastic PMCs, the role the interphase play in key bulk properties of interest to the automotive industry.

  17. Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report

    SciTech Connect (OSTI)

    Swain, A D; Guttmann, H E

    1983-08-01

    The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks.

  18. Abnormal condition and events analysis for instrumentation and control systems. Volume 2: Survey and evaluation of industry practices. Final report

    SciTech Connect (OSTI)

    McKemy, S.; Marcelli, M.; Boehmer, N.; Crandall, D.

    1996-01-01

    The ACES Project was initiated to identify a cost-effective methodology for addressing abnormal conditions and events (ACES) in digital upgrades to nuclear power plant systems, as introduced by IEEE Standard 7-4.3.2-1993. Several methodologies and techniques currently in use in the defense, aerospace, and other communities for the assurance of digital safety systems were surveyed, and although several were shown to possess desirable qualities, none sufficiently met the needs of the nuclear power industry. This report describes a tailorable methodology for performing ACES analysis that is based on the more desirable aspects of the reviewed methodologies and techniques. The methodology is applicable to both safety- and non-safety-grade systems, addresses hardware, software, and system-level concerns, and can be applied in either a lifecycle or post-design timeframe. Employing this methodology for safety systems should facilitate the digital upgrade licensing process.

  19. Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

    SciTech Connect (OSTI)

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Diego Mandelli; Michael Pernice; Robert Nourgaliev

    2013-10-01

    A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis

  20. Energy Engineering Analysis Program - Fort Polk, Louisiana. Executive summary. Final report

    SciTech Connect (OSTI)

    1986-03-01

    The CRS Group, Inc. is pleased to submit this Final Report covering Increments A, B, E, and G of the Energy Engineering Analysis Program (EEAP) for Fort Polk, Louisiana work accomplished under Contract DACA63-80-C-0166 plus modifications with the Fort Worth District, Corps of Engineers. The work presented in these volumes presents the results in which 32 energy conserving measures (ECM`s) have been investigated for ECIP potential for Increments A (existing buildings including Family Housing) and B (existing utilities and energy distribution systems and a centralized Energy Monitoring and Control System-EMCS). Additionally, five scenarios for implementation of a Central Energy Plant (CEP) were studied under Increment E and an analysis of using waste POL as a fuel source in three scenarios was made. Four projects were analyzed under Increment G to identify additional energy savings.

  1. Guidance on risk analysis and safety implications of a large liquefied natural gas (LNG) spill over water.

    SciTech Connect (OSTI)

    Wellman, Gerald William; Melof, Brian Matthew; Luketa-Hanlin, Anay Josephine; Hightower, Marion Michael; Covan, John Morgan; Gritzo, Louis Alan; Irwin, Michael James; Kaneshige, Michael Jiro; Morrow, Charles W.

    2004-12-01

    While recognized standards exist for the systematic safety analysis of potential spills or releases from LNG (Liquefied Natural Gas) storage terminals and facilities on land, no equivalent set of standards or guidance exists for the evaluation of the safety or consequences from LNG spills over water. Heightened security awareness and energy surety issues have increased industry's and the public's attention to these activities. The report reviews several existing studies of LNG spills with respect to their assumptions, inputs, models, and experimental data. Based on this review and further analysis, the report provides guidance on the appropriateness of models, assumptions, and risk management to address public safety and property relative to a potential LNG spill over water.

  2. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  3. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    SciTech Connect (OSTI)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local hot spots do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on first principles. Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of

  4. The SAS4A/SASSYS-1 Safety Analysis Code System

    Energy Science and Technology Software Center (OSTI)

    2012-01-31

    SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal,hydraulic, kinetic, and mechanical phenomena are employed to describe the responsemore » of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolantflow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant

  5. Vehicle technologies heavy vehicle program : FY 2008 benefits analysis, methodology and results --- final report.

    SciTech Connect (OSTI)

    Singh, M.; Energy Systems; TA Engineering

    2008-02-29

    This report describes the approach to estimating the benefits and analysis results for the Heavy Vehicle Technologies activities of the Vehicle Technologies (VT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identifying technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, and (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 08 the Heavy Vehicles program continued its involvement with various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. These changes are the result of a planning effort that first occurred during FY 04 and was updated in the past year. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. The market penetrations are used as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY08 Budget Request. The energy savings models are utilized by the VT program for internal project management purposes.

  6. Safety analysis for tank 241-AZ-101 mixer pump process test

    SciTech Connect (OSTI)

    Milliken, N.J., Westinghouse Hanford

    1996-08-01

    This document establishes the safety envelope for Project W-151,the process test of two mixer pumps in AWF waste tank 241-AZ-101.

  7. Microsoft Word - Final MELCOR Guidance Report Version May 3 2004.doc

    Office of Environmental Management (EM)

    MELCOR Computer Code Application Guidance for Leak Path Factor in Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health U.S. Department of Energy 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 MELCOR LPF Guidance May 2004 Final Report ii INTENTIONALLY BLANK MELCOR LPF Guidance May 2004 Final Report iii Foreword This document provides guidance to Department of Energy (DOE) facility analysts in the use of the MELCOR computer

  8. Enforcement Guidance Supplement 99-03: Limitation of 10 CFR Part 830 to Equipment Referenced in the Safety Analysis Report

    Broader source: Energy.gov [DOE]

    Recently this Office received a reply to a Preliminary Notice of Violation (PNOV), although not denying any facts or conclusions in the PNOV and agreeing to pay the full imposed Civil Penalty, included arguments that some of the equipment cited in the PNOV was not, in their view, subject to the requirements of Part 830. The contractor argued that only equipment referenced in the Safety Analysis Report (SAR), Technical Safety Requirements (TSR) or Technical Specifications should come under the requirements of Part 830. The attached is DOE’s response to denying that argument.

  9. Facility Safety

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

    1996-10-24

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  10. Facility Safety

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

    1995-11-16

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  11. Cutoff walls and cap for lime and M-1 settling basins, Rocky Mountain Arsenal, Colorado. Part 1: Final design analysis. Final report

    SciTech Connect (OSTI)

    1990-10-01

    This document consists of 2 parts, final design analysis and specifications. The purpose of the project was to develop a design for the Interim Response Actions (IRA) at the Lime and M-l Settling Basins at Rocky Mountain Arsenal (RMA), Commerce City, Colorado. The purpose of the IRA at the Lime and M-l Settling Basins is to mitigate the threat of release from the Basins on an interim basis, pending determination of the final remedy in the Onpost Record of Decision (ROD). The IRA for the M-l Basins also includes treatment of the waste materials in the basins with in-situ vitrification (ISV), which is being designed by contract with Woodward-Clyde Consultants.

  12. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    SciTech Connect (OSTI)

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

  13. Safety analysis for the use of hazardous production materials in photovoltaic applications

    SciTech Connect (OSTI)

    Moskowitz, P.D.; Fthenakis, V.M.; Crandall, R.S.; Nelson, B.P.

    1993-12-31

    A wide range of hazardous production materials (HPMs) are used in industrial and university facilities engaged in research and development (R&D) related to semiconductor and photovoltaic devices. Because of the nature of R&D facilities where research activities are constantly changing, it is important for facility managers to pro-actively control the storage, distribution, use and disposal of these HPMs. As part of this control process, facility managers must determine the magnitude of the risk presented by their operations and the protection afforded by the administrative, engineering and personnel controls that have been implemented to reduce risks to life and property to acceptable levels. Facility auditing combined with process hazard analysis (PHA), provides a mechanism for identifying these risks and evaluating their magnitude. In this paper, the methods and results of a PHA for a photovoltaic R&D facility handling HPMs are presented. Of the 30 potential accidents identified, none present High or even Moderate Risks; 18 present Low Risks; and, 12 present Routine Risks. Administrative, engineering and personal safety controls associated with each accident are discussed. 15 refs., 2 figs., 6 tabs.

  14. Safety analysis for the use of hazardous production materials in photovoltaic applications

    SciTech Connect (OSTI)

    Moskowitz, P.D.; Fthenakis, V.M.; Crandall, R.S.; Nelson, B.P.

    1993-11-01

    A wide range of hazardous production materials (HPMs) are used in industrial and university facilities engaged in research and development (R and D) related to semiconductor and photovoltaic devices. Because of the nature of R and D facilities where research activities are constantly changing, it is important for facility managers to pro-actively control the storage, distribution, use and disposal of these HPMs. As part of this control process, facility managers must determine the magnitude of the risk presented by their operations and the protection afforded by the administrative, engineering and personnel controls that have been implemented to reduce risks to life and property to acceptable levels. Facility auditing combined with process hazard analysis (PHA), provides a mechanism for identifying these risks and evaluating their magnitude. In this paper, the methods and results of a PHA for a photovoltaic R and D facility handling HPMs are presented. Of the 30 potential accidents identified, none present High or even Moderate Risks; 18 present Low Risks; and, 12 present Routine Risks. Administrative, engineering and personal safety controls associated with each accident are discussed.

  15. Common problems in the elicitation and analysis of expert opinion affecting probabilistic safety assessments

    SciTech Connect (OSTI)

    Meyer, M.A.; Booker, J.M.

    1990-01-01

    Expert opinion is frequently used in probabilistic safety assessment (PSA), particularly in estimating low probability events. In this paper, we discuss some of the common problems encountered in eliciting and analyzing expert opinion data and offer solutions or recommendations. The problems are: that experts are not naturally Bayesian. People fail to update their existing information to account for new information as it becomes available, as would be predicted by the Bayesian philosophy; that experts cannot be fully calibrated. To calibrate experts, the feedback from the known quantities must be immediate, frequent, and specific to the task; that experts are limited in the number of things that they can mentally juggle at a time to 7 {plus minus} 2; that data gatherers and analysts can introduce bias by unintentionally causing an altering of the expert's thinking or answers; that the level of detail the data, or granularity, can affect the analyses; and the conditioning effect poses difficulties in gathering and analyzing of the expert data. The data that the expert gives can be conditioned on a variety of factors that can affect the analysis and the interpretation of the results. 31 refs.

  16. Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case

    SciTech Connect (OSTI)

    D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

    2014-06-01

    Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

  17. Occupational Safety Performance | Department of Energy

    Office of Environmental Management (EM)

    Occupational Safety Performance Occupational Safety Performance July 19, 2012 Occupational Safety Performance, DOE Quarterly TRC and DART Case Rates - Corporate Analysis of DOE ...

  18. Hydrogen Fuel Cell Analysis: Lessons Learned from Stationary Power Generation Final Report

    SciTech Connect (OSTI)

    Scott E. Grasman; John W. Sheffield; Fatih Dogan; Sunggyu Lee; Umit O. Koylu; Angie Rolufs

    2010-04-30

    This study considered opportunities for hydrogen in stationary applications in order to make recommendations related to RD&D strategies that incorporate lessons learned and best practices from relevant national and international stationary power efforts, as well as cost and environmental modeling of pathways. The study analyzed the different strategies utilized in power generation systems and identified the different challenges and opportunities for producing and using hydrogen as an energy carrier. Specific objectives included both a synopsis/critical analysis of lessons learned from previous stationary power programs and recommendations for a strategy for hydrogen infrastructure deployment. This strategy incorporates all hydrogen pathways and a combination of distributed power generating stations, and provides an overview of stationary power markets, benefits of hydrogen-based stationary power systems, and competitive and technological challenges. The motivation for this project was to identify the lessons learned from prior stationary power programs, including the most significant obstacles, how these obstacles have been approached, outcomes of the programs, and how this information can be used by the Hydrogen, Fuel Cells & Infrastructure Technologies Program to meet program objectives primarily related to hydrogen pathway technologies (production, storage, and delivery) and implementation of fuel cell technologies for distributed stationary power. In addition, the lessons learned address environmental and safety concerns, including codes and standards, and education of key stakeholders.

  19. ANSI/ASHRAE/IES Standard 90.1-2010 Final Determination Quantitative Analysis

    SciTech Connect (OSTI)

    Halverson, Mark A.; Rosenberg, Michael I.; Liu, Bing

    2011-10-31

    The U.S. Department of Energy (DOE) conducted a final quantitative analysis to assess whether buildings constructed according to the requirements of the American National Standards Institute (ANSI)/American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE)/Illuminating Engineering Society of North America (IESNA) Standard 90.1-2010 (ASHRAE Standard 90.1-2010, Standard 90.1-2010, or 2010 edition) would result in energy savings compared with buildings constructed to ANSI/ASHRAE/IESNA Standard 90.1-2007(ASHRAE Standard 90.1-2007, Standard 90.1-2007, or 2007 edition). The final analysis considered each of the 109 addenda to ASHRAE Standard 90.1-2007 that were included in ASHRAE Standard 90.1-2010. All 109 addenda processed by ASHRAE in the creation of Standard 90.1-2010 from Standard 90.1-2007 were reviewed by DOE, and their combined impact on a suite of 16 building prototype models in 15 ASHRAE climate zones was considered. Most addenda were deemed to have little quantifiable impact on building efficiency for the purpose of DOE's final determination. However, out of the 109 addenda, 34 were preliminarily determined to have a measureable and quantifiable impact. A suite of 240 computer energy simulations for building prototypes complying with ASHRAE 90.1-2007 was developed. These prototypes were then modified in accordance with these 34 addenda to create a second suite of corresponding building simulations reflecting the same buildings compliant with Standard 90.1-2010. The building simulations were conducted using the DOE EnergyPlus building simulation software. The resulting energy use from the complete suite of 480 simulation runs was then converted to energy use intensity (EUI, or energy use per unit floor area) metrics (Site EUI, Primary EUI, and energy cost intensity [ECI]) results for each simulation. For each edition of the standard, these EUIs were then aggregated to a national basis for each prototype using weighting factors based on

  20. Selected Area Fishery Evaluation Project Economic Analysis Study Final Report, Final Draft Revision 4: November 10, 2006.

    SciTech Connect (OSTI)

    Bonneville Power Administration; Washington Department of Fish and Wildlife; Oregon Department of Fish and Wildlife

    2006-11-01

    The purpose of this Study is to provide an economic review of current and proposed changes to the Select Area Fishery Evaluation Project (SAFE or Project). The Study results are the information requested in comments made on the Project by a joint review dated March 2005 by the Northwest Power and Conservation Council (NPCC) Independent Scientific Review Panel (ISRP) and Independent Economic Analysis Board (IEAB). North et al. (2006) addressed technical questions about operations and plans, and this report contains the response information for comments concerning Project economics. This report can be considered an economic feasibility review meeting guidelines for cost-effective analysis developed by the IEAB (2003). It also contains other economic measurement descriptions to illustrate the economic effects of SAFE. The SAFE is an expansion of a hatchery project (locally called the Clatsop Economic Development Council Fisheries Project or CEDC) started in 1977 that released an early run coho (COH) stock into the Youngs River. The Youngs River entrance to the Columbia River at River Mile 12 is called Youngs Bay, which is located near Astoria, Oregon. The purpose of the hatchery project was to provide increased fishing opportunities for the in-river commercial fishing gillnet fleet. Instead of just releasing fish at the hatchery, a small scale net pen acclimation project in Youngs Bay was tried in 1987. Hirose et al. (1998) found that 1991-1992 COH broodstock over-wintered at the net pens had double the smolt-to-adult return rate (SAR) of traditional hatchery release, less than one percent stray rates, and 99 percent fishery harvests. It was surmised that smolts from other Columbia River hatcheries could be hauled to the net pens for acclimation and release to take advantage of the SAR's and fishing rates. Proposals were tendered to Bonneville Power Administration (BPA) and other agencies to fund the expansion for using other hatcheries smolts and other off

  1. Nuclear Safety Regulatory Framework

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

    Department of Energy Nuclear Safety Regulatory Framework DOE's Nuclear Safety Enabling Legislation Regulatory Enforcement & Oversight Regulatory Governance Atomic Energy Act 1946 Atomic Energy Act 1954 Energy Reorganization Act 1974 DOE Act 1977 Authority and responsibility to regulate nuclear safety at DOE facilities 10 CFR 830 10 CFR 835 10 CFR 820 Regulatory Implementation Nuclear Safety Radiological Safety Procedural Rules ISMS-QA; Operating Experience; Metrics and Analysis Cross Cutting

  2. Environment/Health/Safety (EHS): Personal Protective Equipment...

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

    EHS Occupational Safety Safety Group Home Electrical Safety Ergonomics ISM Occupational Safety Group Organization Personal Protective Equipment (PPE) Injury Review & Analysis...

  3. Design of Roadside Safety Features

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

    Analysis and Design of the Roadside Safety Features for Safety Performance Texas Transportation Institute (TTI) researchers are investigating the performance of a crash wall design ...

  4. Webinar: An Analysis of AWEA’s Safety Data Report 2015

    Broader source: Energy.gov [DOE]

    The American Wind Energy Association (AWEA) collects safety data in the wind industry and produces a report annually. The intent of the report is to help companies benchmark themselves against...

  5. Statistical Analysis of Occupational Safety Data of Voluntary Protection Program (VPP) and Non-VPP Sites

    Broader source: Energy.gov [DOE]

    The Voluntary Protection Program (VPP) was originally developed by Occupational Safety and Health Administration (OSHA) in 1982 to foster greater ownership of safety and health in the workplace. The Department of Energy (DOE) adopted VPP in 1992; currently 23 sites across the DOE complex participate in the program. As its name implies, it is a voluntary program; i.e. not required by laws or regulations.

  6. Canister storage building hazard analysis report

    SciTech Connect (OSTI)

    Krahn, D.E.; Garvin, L.J.

    1997-07-01

    This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the final CSB safety analysis report (SAR) and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Report, and implements the requirements of DOE Order 5480.23, Nuclear Safety Analysis Report.

  7. Radiation Safety

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

    Safety Home MSDS Search MSDS Help Safety Training and Tests Contact Links LSU Campus Safety Glossary Radiation Safety Manual Radiation Safety Test NOTE: All Training and Testing Material is for LSU CAMD Users ONLY! **Please allow two weeks for your badge to be processed.** Regulations and Hierarchy The CAMD Safety Officer reports to two separate individuals regarding safety. These are the Radiation Safety Officer for the University, and the Campus Safety Officer in all other matters. Thus safety

  8. Engineering Evaluation/Cost Analysis for Power Burst Facility (PER-620) Final End State and PBF Vessel Disposal

    SciTech Connect (OSTI)

    B. C. Culp

    2007-05-01

    Preparation of this engineering evaluation/cost analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, (DOE and EPA 1995) which establishes the Comprehensive Environmental, Response, Compensation, and Liability Act non-time critical removal action process as an approach for decommissioning. The scope of this engineering evaluation/cost analysis is to evaluate alternatives and recommend a preferred alternative for the final end state of the PBF and the final disposal location for the PBF vessel.

  9. Statistical analysis of an inter-laboratory comparison of small-scale safety and thermal testing of RDX

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

    Brown, Geoffrey W.; Sandstrom, Mary M.; Preston, Daniel N.; Pollard, Colin J.; Warner, Kirstin F.; Sorensen, Daniel N.; Remmers, Daniel L.; Phillips, Jason J.; Shelley, Timothy J.; Reyes, Jose A.; et al

    2014-11-17

    In this study, the Integrated Data Collection Analysis (IDCA) program has conducted a proficiency test for small-scale safety and thermal (SSST) testing of homemade explosives (HMEs). Described here are statistical analyses of the results from this test for impact, friction, electrostatic discharge, and differential scanning calorimetry analysis of the RDX Class 5 Type II standard. The material was tested as a well-characterized standard several times during the proficiency test to assess differences among participants and the range of results that may arise for well-behaved explosive materials.

  10. Final Meeting Summary Page 1

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

    ... administrative controls, worker training, industrial hygiene procedures, occupational medicine programs and oversight 2009-present" Final Meeting Summary Page 4 Health, Safety ...

  11. Analysis of the safety and pharmacodynamics of human fibrinogen concentrate in animals

    SciTech Connect (OSTI)

    Beyerle, Andrea; Nolte, Marc W.; Solomon, Cristina; Herzog, Eva; Dickneite, Gerhard

    2014-10-01

    Fibrinogen, a soluble 340 kDa plasma glycoprotein, is critical in achieving and maintaining hemostasis. Reduced fibrinogen levels are associated with an increased risk of bleeding and recent research has investigated the efficacy of fibrinogen concentrate for controlling perioperative bleeding. European guidelines on the management of perioperative bleeding recommend the use of fibrinogen concentrate if significant bleeding is accompanied by plasma fibrinogen levels less than 1.5–2.0 g/l. Plasma-derived human fibrinogen concentrate has been available for therapeutic use since 1956. The overall aim of the comprehensive series of non-clinical investigations presented was to evaluate i) the pharmacodynamic and pharmacokinetic characteristics and ii) the safety and tolerability profile of human fibrinogen concentrate Haemocomplettan P® (RiaSTAP®). Pharmacodynamic characteristics were assessed in rabbits, pharmacokinetic parameters were determined in rabbits and rats and a safety pharmacology study was performed in beagle dogs. Additional toxicology tests included: single-dose toxicity tests in mice and rats; local tolerance tests in rabbits; and neoantigenicity tests in rabbits and guinea pigs following the introduction of pasteurization in the manufacturing process. Human fibrinogen concentrate was shown to be pharmacodynamically active in rabbits and dogs and well tolerated, with no adverse events and no influence on circulation, respiration or hematological parameters in rabbits, mice, rats and dogs. In these non-clinical investigations, human fibrinogen concentrate showed a good safety profile. This data adds to the safety information available to date, strengthening the current body of knowledge regarding this hemostatic agent. - Highlights: • A comprehensive series of pre-clinical investigations of human fibrinogen concentrate. • Human fibrinogen concentrate was shown to be pharmacodynamically active. • Human fibrinogen concentrate was well tolerated

  12. Guide for reviewing safety analysis reports for packaging: Review of quality assurance requirements

    SciTech Connect (OSTI)

    Moon, D.W.

    1988-10-01

    This review section describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety. The design effort, operation's plans, and quality assurance requirements should be integrated to achieve a system in which the independent QA program is not overly stringent and the application of QA requirements is commensurate with safety significance. The reviewer must verify that the applicant's QA section in the SARP contains package-specific QA information required by DOE Orders and federal regulations that demonstrate compliance. 8 refs.

  13. TWRS safety management plan

    SciTech Connect (OSTI)

    Popielarczyk, R.S., Westinghouse Hanford

    1996-08-01

    The Tank Waste Remediation System (TWRS) Safety Management Program Plan for development, implementation and maintenance of the tank farm authorization basis is described. The plan includes activities and procedures for: (a) Updating the current Interim Safety Basis, (b) Development,implementation and maintenance of a Basis for Interim Operations, (c) Development, implementation and maintenance of the Final Safety Analyses Report, (d) Development and implementation of a TWRS information Management System for monitoring the authorization basis.

  14. Status Update on Action 1b: Analysis of WP&C Deficiencies Identified by the DNFSB

    Broader source: Energy.gov [DOE]

    Slide Presentation by Stephen L. Domotor, Office of Analysis, Office of Health, Safety and Security. Office of Analysis, Office of Health, Safety and Security. Analysis of Integrated Safety Management at the Activity Level: Work Planning and Control-Final Report,, U.S. Department of Energy, August 1, 2013.

  15. Job Safety

    Office of Environmental Management (EM)

    Job Safety and Health It's the law EMPLOYEES: Must have access to: DOE safety and health publications; The worker safety and health program for their location; This ...

  16. Summary final report: Contract between the Japan atomic power company and the U.S. Department of Energy Improvement of core safety - study on GEM (III)

    SciTech Connect (OSTI)

    Burke, T.M.; Lucoff, D.M.

    1997-03-18

    This report provides a summary of activities associated with the technical exchange between representatives of the Japan Atomic Power Company (JAPC) and the United States Department of Energy (DOE) regarding the development and testing of Gas Expansion Modules (GEM) at the Fast Flux Test Facility (FFTF). Issuance of this report completes the scope of work defined in the original contract between JAPC and DOE titled ''Study on Improvement of Core Safety - Study on GEM (III).'' Negotiations related to potential modification of the contract are in progress. Under the proposed contract modification, DOE would provide an additional report documenting FFTF pump start tests with GEMs and answer additional JAPC questions related to core safety with and without GEMs.

  17. Analysis of safety precautions for coal and gas outburst-hazardous strata

    SciTech Connect (OSTI)

    Hudecek, V.

    2008-09-15

    The author analyses coal and gas outbursts and generalizes the available data on the approaches to solving the problematics of these gas-dynamic events in the framework of Czech Republic Grant 'Estimate of the Safety Precautions for Coal and Gas Outburst Hazardous Strata'.

  18. Application of a Comprehensive Sensitivity Analysis Method on the Safety Assessment of TRU Waste Disposal in Japan

    SciTech Connect (OSTI)

    Takao Ohi; Manabu Inagaki; Tomoyuki Sone; Morihiro Mihara; Takeshi Ebashi; Hiroyasu Takase; Kiyoshi Oyamada; Kunihiko Nakajima

    2007-07-01

    A comprehensive sensitivity analysis method has been developed with the aim of providing quantitative information in an efficient manner. This methodology is composed of the following two components: (1) a statistical method with random sampling of independent parameters, which identifies important parameters and extracts threshold values of parameters and/or combinations yielding a 'successful condition' where maximum dose does not exceed a target value, (2) A nuclide migration model that as far as possible incorporates a comprehensive set of phenomena occurring within the repository. This approach was applied as part of a safety assessment of the geological disposal of transuranic (TRU) waste in Japan (TRU-2). It was shown that the concept of TRU waste disposal is robust from the point of view of safety. (authors)

  19. Hawaii demand-side management resource assessment. Final report, Reference Volume 1: Building prototype analysis

    SciTech Connect (OSTI)

    1995-04-01

    This report provides a detailed description of, and the baseline assumptions and simulation results for, the building prototype simulations conducted for the building types designated in the Work Plan for Demand-side Management Assessment of Hawaii`s Demand-Side Resources (HES-4, Phase 2). This report represents the second revision to the initial building prototype description report provided to DBEDT early in the project. Modifications and revisions to the prototypes, based on further calibration efforts and on comments received from DBEDT Staff have been incorporated into this final version. These baseline prototypes form the basis upon which the DSM measure impact estimates and the DSM measure data base were developed for this project. This report presents detailed information for each of the 17 different building prototypes developed for use with the DOE-21E program (23 buildings in total, including resorts and hotels defined separately for each island) to estimate the impact of the building technologies and measures included in this project. The remainder of this section presents some nomenclature and terminology utilized in the reports, tables, and data bases developed from this project to denote building type and vintage. Section 2 contains a more detailed discussion of the data sources, the definition of the residential sector building prototypes, and results of the DOE-2 analysis. Section 3 provides a similar discussion for the commercial sector. The prototype and baseline simulation results are presented in a separate section for each building type. Where possible, comparison of the baseline simulation results with benchmark data from the ENERGY 2020 model or other demand forecasting models specific to Hawaii is included for each building. Appendix A contains a detailed listing of the commercial sector baseline indoor lighting technologies included in the existing and new prototypes by building type.

  20. Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689Nuclear Energy Institute (NEI) Letter, 9/10/10

    Broader source: Energy.gov [DOE]

    Enclosed for your review is a Nuclear Energy Institute white paper on the use of Integrated Safety Analysis (ISA) at U.S. Nuclear Regulatory Commission-licensed recycling facilities. This paper is...

  1. CRAD, Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities

    Office of Energy Efficiency and Renewable Energy (EERE)

    These guidelines and criteria provide a consistent overall framework for assessment of the processes that are currently in place to ensure that the software being used in the safety analysis and design of the SSCs in defense nuclear facilities is adequate. These reviews will be conducted only on software that is currently in use, not on software that was previously used as part of a safety analysis and design process.

  2. Facility Safety

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

    2013-06-21

    DOE-STD-1104 contains the Department's method and criteria for reviewing and approving nuclear facility's documented safety analysis (DSA). This review and approval formally document the basis for DOE, concluding that a facility can be operated safely in a manner that adequately protects workers, the public, and the environment. Therefore, it is appropriate to formally require implementation of the review methodology and criteria contained in DOE-STD-1104.

  3. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    SciTech Connect (OSTI)

    Enercon Services, Inc.

    2011-03-14

    ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost

  4. Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory TRU Californium Shipping Container

    SciTech Connect (OSTI)

    Box, W.D.; Shappert, L.B.; Seagren, R.D.; Klima, B.B.; Jurgensen, M.C.; Hammond, C.R.; Watson, C.D.

    1980-01-01

    An analytical evaluation of the Oak Ridge National Laboratory TRU Californium Shipping Container was made in order to demonstrate its compliance with the regulations governing off-site shipment of packages that contain radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of this evaluation demonstrate that the container complies with the applicable regulations.

  5. Packaging review guide for reviewing safety analysis reports for packagings: Revision 1

    SciTech Connect (OSTI)

    Fisher, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

    1988-10-01

    The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. The principal purpose of this document is to assure the quality and uniformity of PCS reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The Packaging Review Guide (PRG) also sets forth solutions and approaches determined to be acceptable in the past in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented. It is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 77 refs., 16 figs., 15 tabs.

  6. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    SciTech Connect (OSTI)

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  7. Cold Vacuum Drying (CVD) Facility Hazards Analysis Report

    SciTech Connect (OSTI)

    CROWE, R.D.

    2000-08-07

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) Hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  8. Final Report for Geometric Analysis for Data Reduction and Structure Discovery DE-FG02-10ER25983, STRIPES award # DE-SC0004096

    SciTech Connect (OSTI)

    Vixie, Kevin R

    2014-11-27

    This is the final report for the project "Geometric Analysis for Data Reduction and Structure Discovery" in which insights and tools from geometric analysis were developed and exploited for their potential to large scale data challenges.

  9. Safety analysis--200 Area Savannah River Site: Separations Area operations Building 211-H Outside Facilities. Supplement 11, Revision 1

    SciTech Connect (OSTI)

    Not Available

    1993-01-01

    The H-Area Outside Facilities are located in the 200-H Separations Area and are comprised of a number of processes, utilities, and services that support the separations function. Included are enriched uranium loadout, bulk chemical storage, water handling, acid recovery, general purpose evaporation, and segregated solvent facilities. In addition, services for water, electricity, and steam are provided. This Safety Analysis Report (SAR) documents an analysis of the H-Area Outside Facilities and is one of a series of documents for the Separations Area as specified in the SR Implementation Plan for DOE order 5481.1A. The primary purpose of the analysis was to demonstrate that the facility can be operated without undue risk to onsite or offsite populations, to the environment, and to operating personnel. In this report, risks are defined as the expected frequencies of accidents, multiplied by the resulting radiological consequences in person-rem. Following the summary description of facility and operations is the site evaluation including the unique features of the H-Area Outside Facilities. The facility and process design are described in Chapter 3.0 and a description of operations and their impact is given in Chapter 4.0. The accident analysis in Chapter 5.0 is followed by a list of safety related structures and systems (Chapter 6.0) and a description of the Quality Assurance program (Chapter 7.0). The accident analysis in this report focuses on estimating the risk from accidents as a result of operation of the facilities. The operations were evaluated on the basis of three considerations: potential radiological hazards, potential chemical toxicity hazards, and potential conditions uniquely different from normal industrial practice.

  10. Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

    SciTech Connect (OSTI)

    Chopra, Omesh K.; Diercks, Dwight R.; Ma, David Chia-Chiun; Garud, Yogendra S.

    2013-12-17

    At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC’s assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978

  11. Safety analysis report for the use of hazardous production materials in photovoltaic applications at the National Renewable Energy Laboratory

    SciTech Connect (OSTI)

    Crandall, R.S.; Nelson, B.P.; Moskowitz, P.D.; Fthenakis, V.M.

    1992-07-01

    To ensure the continued safety of SERI's employees, the community, and the environment, NREL commissioned an internal audit of its photovoltaic operations that used hazardous production materials (HPMS). As a result of this audit, NREL management voluntarily suspended all operations using toxic and/or pyrophoric gases. This suspension affected seven laboratories and ten individual deposition systems. These activities are located in Building 16, which has a permitted occupancy of Group B, Division 2 (B-2). NREL management decided to do the following. (1) Exclude from this SAR all operations which conformed, or could easily be made to conform, to B-2 Occupancy requirements. (2) Include in this SAR all operations that could be made to conform to B-2 Occupancy requirements with special administrative and engineering controls. (3) Move all operations that could not practically be made to conform to B-2 occupancy requirements to alternate locations. In addition to the layered set of administrative and engineering controls set forth in this SAR, a semiquantitative risk analysis was performed on 30 various accident scenarios. Twelve presented only routine risks, while 18 presented low risks. Considering the demonstrated safe operating history of NREL in general and these systems specifically, the nature of the risks identified, and the layered set of administrative and engineering controls, it is clear that this facility falls within the DOE Low Hazard Class. Each operation can restart only after it has passed an Operational Readiness Review, comparing it to the requirements of this SAR, while subsequent safety inspections will ensure future compliance. This document contains the appendices to the NREL safety analysis report.

  12. Analysis of FERC's Final EIS for Electricity Open Access & Recovery of Stranded Costs

    Reports and Publications (EIA)

    1996-01-01

    Reviews the Final Environmental Impact Statement (FEIS) prepared by the Federal Energy Regulatory Commission for its electricity transmission system open access prepared in April 1996 and uses the National Energy Modeling System (NEMS) to analyze the open access rule (Orders 888 and 889).

  13. WIPP Documents - Nuclear Safety

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

    Nuclear Safety DOE/WIPP-07-3372, Revision 5b, WIPP Documented Safety Analysis Approved April 2016 The Documented Safety Analysis addresses all hazards (both radiological and nonradiological) and the controls necessary to provide adequate protection to the public, workers, and the environment. The WIPP DSA demonstrates the extent to which the Waste Isolation Pilot Plant can be operated safely with respect to workers, the public, and the environment. DOE/WIPP-07-3373, Revision 5b, WIPP Technical

  14. DOE's Safety Bulletin No. 2011-01, Events Beyond Design Safety Basis

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

    Analysis, March 2011 | Department of Energy DOE's Safety Bulletin No. 2011-01, Events Beyond Design Safety Basis Analysis, March 2011 DOE's Safety Bulletin No. 2011-01, Events Beyond Design Safety Basis Analysis, March 2011 PURPOSE This Safety Alert provides information on a safety concern related to the identification and mitigation of events that may fall outside those analyzed in the documented safety analysis. BACKGROUND On March 11, 2011, the Fukushima Daiichi nuclear power station in

  15. An OSHA based approach to safety analysis for nonradiological hazardous materials

    SciTech Connect (OSTI)

    Yurconic, M.

    1992-08-01

    The PNL method for chemical hazard classification defines major hazards by means of a list of hazardous substances (or chemical groups) with associated trigger quantities. In addition, the functional characteristics of the facility being classified is also be factored into the classification. In this way, installations defined as major hazard will only be those which have the potential for causing very serious incidents both on and off site. Because of the diversity of operations involving chemicals, it may not be possible to restrict major hazard facilities to certain types of operations. However, this hazard classification method recognizes that in the industrial sector major hazards are most commonly associated with activities involving very large quantities of chemicals and inherently energetic processes. These include operations like petrochemical plants, chemical production, LPG storage, explosives manufacturing, and facilities which use chlorine, ammonia, or other highly toxic gases in bulk quantities. The basis for this methodology is derived from concepts used by OSHA in its proposed chemical process safety standard, the Dow Fire and Explosion Index Hazard Classification Guide, and the International Labor Office`s program on chemical safety. For the purpose of identifying major hazard facilities, this method uses two sorting criteria, (1) facility function and processes and (2) quantity of substances to identify facilities requiringclassification. Then, a measure of chemical energy potential (material factor) is used to identify high hazard class facilities.

  16. An OSHA based approach to safety analysis for nonradiological hazardous materials

    SciTech Connect (OSTI)

    Yurconic, M.

    1992-08-01

    The PNL method for chemical hazard classification defines major hazards by means of a list of hazardous substances (or chemical groups) with associated trigger quantities. In addition, the functional characteristics of the facility being classified is also be factored into the classification. In this way, installations defined as major hazard will only be those which have the potential for causing very serious incidents both on and off site. Because of the diversity of operations involving chemicals, it may not be possible to restrict major hazard facilities to certain types of operations. However, this hazard classification method recognizes that in the industrial sector major hazards are most commonly associated with activities involving very large quantities of chemicals and inherently energetic processes. These include operations like petrochemical plants, chemical production, LPG storage, explosives manufacturing, and facilities which use chlorine, ammonia, or other highly toxic gases in bulk quantities. The basis for this methodology is derived from concepts used by OSHA in its proposed chemical process safety standard, the Dow Fire and Explosion Index Hazard Classification Guide, and the International Labor Office's program on chemical safety. For the purpose of identifying major hazard facilities, this method uses two sorting criteria, (1) facility function and processes and (2) quantity of substances to identify facilities requiringclassification. Then, a measure of chemical energy potential (material factor) is used to identify high hazard class facilities.

  17. Statistical Analysis of the Worker Engagement Survey Administered at the Worker Safety and Security Team Festival

    SciTech Connect (OSTI)

    Davis, Adam Christopher

    2015-08-25

    The Worker Safety and Security Team (WSST) at Los Alamos National Laboratory holds an annual festival, WSST-fest, to engage workers and inform them about safety- and securityrelated matters. As part of the 2015 WSST-fest, workers were given the opportunity to participate in a survey assessing their engagement in their organizations and work environments. A total of 789 workers participated in the 23-question survey where they were also invited, optionally, to identify themselves, their organization, and to give open-ended feedback. The survey consisted of 23 positive statements (i.e. “My organization is a good place to work.”) with which the respondent could express a level of agreement. The text of these statements are provided in Table 1. The level of agreement corresponds to a 5-level Likert scale ranging from “Strongly Disagree” to “Strongly Agree.” In addition to assessing the overall positivity or negativity of the scores, the results were partitioned into several cohorts based on the response meta-data (self-identification, comments, etc.) to explore trends. Survey respondents were presented with the options to identify themselves, their organizations and to provide comments. These options suggested the following questions about the data set.

  18. Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

    SciTech Connect (OSTI)

    OSCAR,DEBBY S.; WALKER,SHARON ANN; HUNTER,REGINA LEE; WALKER,CHERYL A.

    1999-12-01

    The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2.

  19. Analysis of Coconut-Derived Biodiesel and Conventional Diesel Fuel Samples from the Philippines: Task 2 Final Report

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

    Analysis of Coconut-Derived Biodiesel and Conventional Diesel Fuel Samples from the Philippines Task 2 Final Report T.L. Alleman and R.L. McCormick Milestone Report NREL/MP-540-38643 January 2006 National Renewable Energy Laboratory 1617 Cole Boulevard, Golden, Colorado 80401-3393 303-275-3000 * www.nrel.gov Operated for the U.S. Department of Energy Office of Energy Efficiency and Renewable Energy by Midwest Research Institute * Battelle Contract No. DE-AC36-99-GO10337 Analysis of Coconut-

  20. Plutonium Finishing Plant safety evaluation report

    SciTech Connect (OSTI)

    Not Available

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91.

  1. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    SciTech Connect (OSTI)

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zubers hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  2. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    SciTech Connect (OSTI)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  3. 2010-2025 Scenario Analysis for Hydrogen Fuel Cell Vehicles and Infrastructure Final List of Attendees

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

    Review and Discussion of Preliminary Results August 9-10, 2006 Omni Shoreham Hotel Washington DC 1 Final List of Attendees Arlene Anderson U.S. Department of Energy Email: arlene.anderson@ee.doe.gov Howard Andres Energetics Incorporated Email: handres@energetics.com Klaus Bonhoff DaimlerChrysler AG Email: klaus.bonhoff@daimlerchrysler.com Brian Bonner Air Products Email: bonnerbb@airproducts.com Chris Bordeaux Air Liquide (Consultant) Ethan Brown Ballard Email: ethan.brown@ballard.com Kwontae

  4. A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

    SciTech Connect (OSTI)

    Disney, R.K.

    1994-10-01

    The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE`s) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE`s within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ``site`` perception to a more uniform or ``national`` perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation.

  5. Radioactive Solid Waste Storage and Disposal at Oak Ridge National Laboratory, Description and Safety Analysis

    SciTech Connect (OSTI)

    Bates, L.D.

    2001-01-30

    Oak Ridge National Laboratory (ORNL) is a principle Department of Energy (DOE) Research Institution operated by the Union Carbide Corporation - Nuclear Division (UCC-ND) under direction of the DOE Oak Ridge Operations Office (DOE-ORO). The Laboratory was established in east Tennessee, near what is now the city of Oak Ridge, in the mid 1940s as a part of the World War II effort to develop a nuclear weapon. Since its inception, disposal of radioactively contaminated materials, both solid and liquid, has been an integral part of Laboratory operations. The purpose of this document is to provide a detailed description of the ORNL Solid Waste Storage Areas, to describe the practice and procedure of their operation, and to address the health and safety impacts and concerns of that operation.

  6. SAS4A/SASSYS-1: Fast Reactor Safety Analysis Code | Argonne National...

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

    The original code development was for sodium-cooled fast reactors, and sodium boiling can be modeled. However, basic core thermal-hydraulics and systems analysis features are ...

  7. Health & Safety

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

    Health & Safety Health & Safety1354608000000Health & SafetySome of these resources are LANL-only and will require Remote Access.NoQuestions? 667-5809library@lanl.gov Health &...

  8. Safety, Security

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

    Safety, Security Safety, Security The Lab's mission is to develop and apply science and technology to ensure the safety, security, and reliability of the U.S. nuclear deterrent; reduce global threats; and solve other emerging national security and energy challenges. Contact Operator Los Alamos National Laboratory (505) 667-5061 We do not compromise safety for personal, programmatic, or operational reasons. Safety: we integrate safety, security, and environmental concerns into every step of our

  9. Safety Analysis Report for the use of hazardous production materials in photovoltaic applications at the National Renewable Energy Laboratory

    SciTech Connect (OSTI)

    Crandall, R.S.; Nelson, B.P. ); Moskowitz, P.D.; Fthenakis, V.M. )

    1992-07-01

    To ensure the continued safety of SERI's employees, the community, and the environment, NREL commissioned an internal audit of its photovoltaic operations that used hazardous production materials (HPMs). As a result of this audit, NREL management voluntarily suspended all operations using toxic and/or pyrophoric gases. This suspension affected seven laboratories and ten individual deposition systems. These activities are located in Building 16, which has a permitted occupancy of Group B, Division 2 (B-2). NREL management decided to do the following. (1) Exclude from this SAR all operations which conformed, or could easily be made to conform, to B-2 Occupancy requirements. (2) Include in this SAR all operations that could be made to conform to B-2 Occupancy requirements with special administrative and engineering controls. (3) Move all operations that could not practically be made to conform to B-2 Occupancy requirements to alternate locations. In addition to the layered set of administrative and engineering controls set forth in this SAR, a semiquantitative risk analysis was performed on 30 various accident scenarios. Twelve presented only routine risks, while 18 presented low risks. Considering the demonstrated safe operating history of NREL in general and these systems specifically, the nature of the risks identified, and the layered set of administrative and engineering controls, it is clear that this facility falls within the DOE Low Hazard Class. Each operation can restart only after it has passed an Operational Readiness Review, comparing it to the requirements of this SAR, while subsequent safety inspections will ensure future compliance.

  10. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    SciTech Connect (OSTI)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-30

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 10{sup 6} cm{sup −1}) in a tube, their delta

  11. 2014 Annual Workforce Analysis and Staffing Plan Report- Office of Environment, Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  12. 2010 Annual Workforce Analysis and Staffing Plan Report- Office of Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  13. 2015 Annual Workforce Analysis and Staffing Plan Report- Office of Environment, Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  14. 2015 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Office of Energy Efficiency and Renewable Energy (EERE)

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  15. 2011 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  16. 2013 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  17. 2010 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  18. 2012 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  19. Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and Example Sheet)

    Broader source: Energy.gov [DOE]

    The Staffing Analysis calculation is completed using an Excel worksheet. Information locations are identified by titles in column or row headings and worksheet locations based on the unmodified blank worksheet attached.

  20. 2013 Annual Workforce Analysis and Staffing Plan Report- Office of Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  1. 2011 Annual Workforce Analysis and Staffing Plan Report- Office of Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  2. 2012 Annual Workforce Analysis and Staffing Plan Report- Office of Health, Safety and Security

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  3. 2014 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  4. CSB Investigations and Safety Culture | Department of Energy

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

    from occurring again. Analysis of Safety Systems CSB Investigations and Safety Culture More Documents & Publications Nuclear Safety Workshop Summary Operating Experience...

  5. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect (OSTI)

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  6. An advanced deterministic method for spent-fuel criticality safety analysis

    SciTech Connect (OSTI)

    DeHart, M.D.

    1998-09-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, nonorthogonal configurations or fissile materials, typical of real-world problems. In the last few years, however, interest in determinist transport methods has been revived, due to shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple-pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constrains in finite differencing schemes have made discrete ordinates methods impractical for nonorthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitation of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built on the ESC formalism, is being developed as part of the SCALE code system. This paper demonstrates the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

  7. Implementation of Revision 19 of the TRUPACT-II Safety Analysis Report at Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    D'Amico, E.; O'Leary, J.; Bell, S.; Djordjevic, S.; Givens, C,; Shokes, T.; Thompson, S.; Stahl, S.

    2003-02-25

    The U.S. Nuclear Regulatory Commission on July 27, 2001 approved Revision 19 of the TRUPACT-II Safety Analysis Report (SAR) and the associated TRUPACT-II Authorized Methods for Payload Control (TRAMPAC). Key initiatives in Revision 19 included matrix depletion, unlimited mixing of shipping categories, a flammability assessment methodology, and an alternative methodology for the determination of flammable gas generation rates. All U.S. Department of Energy (DOE) sites shipping transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) were required to implement Revision 19 methodology into their characterization and waste transportation programs by May 20, 2002. An implementation process was demonstrated by the Rocky Flats Environmental Technology Site (RFETS) in Golden, Colorado. The three-part process used by RFETS included revision of the site-specific TRAMPAC, an evaluation of the contact-handled TRU waste inventory against the regulations in Revision 19, and design and development of software to facilitate future inventory analyses.

  8. On the Partial-Wave Analysis of Mesonic Resonances Decaying to Multiparticle Final States Produced by Polarized Photons

    SciTech Connect (OSTI)

    Salgado, Carlos W.; Weygand, Dennis P.

    2014-04-01

    Meson spectroscopy is going through a revival with the advent of high statistics experiments and new advances in the theoretical predictions. The Constituent Quark Model (CQM) is finally being expanded considering more basic principles of field theory and using discrete calculations of Quantum Chromodynamics (lattice QCD). These new calculations are approaching predictive power for the spectrum of hadronic resonances and decay modes. It will be the task of the new experiments to extract the meson spectrum from the data and compare with those predictions. The goal of this report is to describe one particular technique for extracting resonance information from multiparticle final states. The technique described here, partial wave analysis based on the helicity formalism, has been used at Brookhaven National Laboratory (BNL) using pion beams, and Jefferson Laboratory (Jlab) using photon beams. In particular this report broaden this technique to include production experiments using linearly polarized real photons or quasi-real photons. This article is of a didactical nature. We describe the process of analysis, detailing assumptions and formalisms, and is directed towards people interested in starting partial wave analysis.

  9. Microsoft Word - Threat Analysis Framework Sept07_comments-final.doc

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

    5792 Unlimited Release September 2007 Threat Analysis Framework David P. Duggan and John T. Michalski Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy's National Nuclear Security Administration under Contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. Threat Analysis

  10. Next Generation Electromagnetic Pump Analysis Tools (PLM DOC-0005-2188). Final Report

    SciTech Connect (OSTI)

    Stregy, Seth; Dasilva, Ana; Yilmaz, Serkan; Saha, Pradip; Loewen, Eric

    2015-10-29

    This report provides the broad historical review of EM Pump development and details of MATRIX development under this project. This report summarizes the efforts made to modernize the legacy performance models used in previous EM Pump designs and the improvements made to the analysis tools. This report provides information on Tasks 1, 3, and 4 of the entire project. The research for Task 4 builds upon Task 1: Update EM Pump Databank and Task 3: Modernize the Existing EM Pump Analysis Model, which are summarized within this report. Where research for Task 2: Insulation Materials Development and Evaluation identified parameters applicable to the analysis model with Task 4, the analysis code was updated, and analyses were made for additional materials. The important design variables for the manufacture and operation of an EM Pump that the model improvement can evaluate are: space constraints; voltage capability of insulation system; maximum flux density through iron; flow rate and outlet pressure; efficiency and manufacturability. The development of the next-generation EM Pump analysis tools during this two-year program provides information in three broad areas: Status of analysis model development; Improvements made to older simulations; and Comparison to experimental data.

  11. Onshore permitting systems analysis for coal, oil, gas, geothermal and oil shale leases. Final report

    SciTech Connect (OSTI)

    Not Available

    1982-09-01

    The magnitude and complexity of permit processes raises a question as to their impact on the rate and scope of industrial development activity. One particular area where this issue is of concern is in new energy extraction and development activities. The initiation of new energy projects has been a national priority for several years. But, energy projects, because of their potential for creating land disturbances, are subject to many environmental and other regulations. Because of this, the permitting required of energy resource developers is extensive. Within the energy field, a major portion of development activities occurs on federal lands. This is particularly true in the Rocky Mountain states and Alaska where the principal landholder is the federal government. The permitting requirements for federal lands' development differ from those for private lands. This report assesses the impact of permitting processes for energy resource development on federal lands. The permitting processes covered include all of the major environmental, land-use, and safety permits required by agencies of federal and state governments. The lands covered include all federal lands, with emphasis on eight states with major development activities.

  12. Site-specific earthquake response analysis for Paducah Gaseous Diffusion Plant, Paducah, Kentucky. Final report

    SciTech Connect (OSTI)

    Sykora, D.W.; Davis, J.J.

    1993-08-01

    The Paducah Gaseous Diffusion Plant (PGDP), owned by the US Department of Energy (DOE) and operated under contract by Martin Marietta Energy systems, Inc., is located southwest of Paducah, Kentucky. An aerial photograph and an oblique sketch of the plant are shown in Figures 1 and 2, respectively. The fenced portion of the plant consists of 748 acres. This plant was constructed in the 1950`s and is one of only two gaseous diffusion plants in operation in the United States; the other is located near Portsmouth, Ohio. The facilities at PGDP are currently being evaluated for safety in response to natural seismic hazards. Design and evaluation guidelines to evaluate the effects of earthquakes and other natural hazards on DOE facilities follow probabilistic hazard models that have been outlined by Kennedy et al. (1990). Criteria also established by Kennedy et al. (1990) classify diffusion plants as ``moderate hazard`` facilities. The US Army Engineer Waterways Experiment Station (WES) was tasked to calculate the site response using site-specific design earthquake records developed by others and the results of previous geotechnical investigations. In all, six earthquake records at three hazard levels and four individual and one average soil columns were used.

  13. Hazard analysis of compressed natural gas fueling systems and fueling procedures used at retail gasoline service stations. Final report

    SciTech Connect (OSTI)

    1995-04-28

    An evaluation of the hazards associated with operations of a typical compressed natural gas (CNG) fueling station is presented. The evaluation includes identification of a typical CNG fueling system; a comparison of the typical system with ANSI/NFPA (American National Standards Institute/National Fire Protection Association) Standard 52, Compressed Natural Gas (CNG) Vehicular Fuel System, requirements; a review of CNG industry safety experience as identified in current literature; hazard identification of potential internal (CNG system-specific causes) and external (interface of co-located causes) events leading to potential accidents; and an analysis of potential accident scenarios as determined from the hazard evaluation. The study considers CNG dispensing equipment and associated equipment, including the compressor station, storate vessels, and fill pressure sensing system.

  14. Thermal reactor safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  15. Electrical Safety Occurrences | Department of Energy

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

    requested by the Electrical Safety Community for information exchange and continual learning. August 16, 2011 Monthly Analysis of Electrical Safety Occurrences - July 2011 An...

  16. Electrical Safety Occurrences | Department of Energy

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

    requested by the Electrical Safety Community for information exchange and continual learning. October 29, 2013 Monthly Analysis of Electrical Safety Occurrences - September 2013...

  17. Abnormal condition and events analysis for instrumentation and control systems. Volume 1: Methodology for nuclear power plant digital upgrades. Final report

    SciTech Connect (OSTI)

    McKemy, S.; Marcelli, M.; Boehmer, N.; Crandall, D.

    1996-01-01

    The ACES project was initiated to identify a cost-effective methodology for addressing abnormal conditions and events (ACES) in digital upgrades to nuclear power plant systems, as introduced by IEEE Standard 7-4.3.2-1993. Several methodologies and techniques currently in use in the defense, aerospace, and other communities for the assurance of digital safety systems were surveyed, and although several were shown to possess desirable qualities, non sufficiently met the needs of the nuclear power industry. This report describes a tailorable methodology for performing ACES analysis that is based on the more desirable aspects of the reviewed methodologies and techniques. The methodology is applicable to both safety- and non-safety-grade systems, addresses hardware, software, and system-level concerns, and can be applied in either a lifecycle or post-design timeframe. Employing this methodology for safety systems should facilitate the digital upgrade licensing process.

  18. A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks

    SciTech Connect (OSTI)

    SMALLEY, J.L.

    1999-05-18

    This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR).

  19. Facility Safety

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

    2000-11-20

    The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.

  20. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect (OSTI)

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  1. Nuclear Safety Policy, Guidance & Reports

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Safety establishes and maintains nuclear safety policy, requirements, and guidance including policy and requirements relating to hazard and accident analysis, facility design and operation, and Quality Assurance.

  2. Energy engineering analysis Iowa Army Ammunition Plant Burlington, Iowa. Final report

    SciTech Connect (OSTI)

    NONE

    1983-05-10

    This Energy Engineering Analysis consists of the main report, three appendices, and a summary of annual energy consumption on a `per building` basis. The main report identifies the purpose of the study, describes the existing and anticipated energy use trends, and defines and summarizes specific energy conservation projects recommended to achieve the goals stated in the Army Facilities Energy Plan. Appendices I, II and III, and the Annual Energy Consumption Summary include building information, weather data, cost data, and detailed computer-generated and manual calculations for each individual project. The analysis will enable ammunition plant personnel to identify energy conservation measures and meet Army energy reduction goals. The report includes: Energy consumption by fuel type Energy consumption trends ECAM projects other potential projects Quick-fix management form Description of analyzed buildings. In addition-, the Analysis is a detailed data base consisting of: An analysis of building energy use Energy Conservation Measures applied to each analyzed building to be improved A set of marked-up prints from the survey indicating the conditions when surveyed.

  3. Statistical planning and analysis for treatments of tar sand wastewater. Final report

    SciTech Connect (OSTI)

    Pirie, W.R.

    1984-03-01

    The first part of this report discusses the overall statistical planning, coordination and design for several tar sand wastewater treatment projects contracted by the Laramie Energy Technology Center (LETC) of the Department of Energy. A general discussion of the benefits of consistent statistical design and analysis for data-oriented projects is included, with recommendations for implementation. A detailed outline of the principles of general linear models design is followed by an introduction to recent developments in general linear models by ranks (GLMR) analysis and a comparison to standard analysis using Gaussian or normal theory (GLMN). A listing of routines contained in the VPI Nonparametric Statistics Package (NPSP), installed on the Cyber computer system at the University of Wyoming is included. Part 2 describes in detail the design and analysis for treatments by Gas Flotation, Foam Separation, Coagulation, and Ozonation, with comparisons among the first three methods. Rank methods are used for most analyses, and several detailed examples are included. For optimization studies, the powerful tools of response surface analysis (RSA) are employed, and several sections contain discussion on the benefits of RSA. All four treatment methods proved to be effective for removal of TOC and suspended solids from the wastewater. Because the processes and equipment designs were new, optimum removals were not achieved by these initial studies and reasons for that are discussed. Pollutant levels were nevertheless reduced to levels appropriate for recycling within the process, and for such reuses as steam generation, according to the DOE/LETC project officer. 12 refs., 8 figs., 21 tabs.

  4. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

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

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-06-15

    Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  5. Final Harvest of Above-Ground Biomass and Allometric Analysis of the Aspen FACE Experiment

    SciTech Connect (OSTI)

    Mark E. Kubiske

    2013-04-15

    The Aspen FACE experiment, located at the US Forest Service Harshaw Research Facility in Oneida County, Wisconsin, exposes the intact canopies of model trembling aspen forests to increased concentrations of atmospheric CO2 and O3. The first full year of treatments was 1998 and final year of elevated CO2 and O3 treatments is scheduled for 2009. This proposal is to conduct an intensive, analytical harvest of the above-ground parts of 24 trees from each of the 12, 30 m diameter treatment plots (total of 288 trees) during June, July & August 2009. This above-ground harvest will be carefully coordinated with the below-ground harvest proposed by D.F. Karnosky et al. (2008 proposal to DOE). We propose to dissect harvested trees according to annual height growth increment and organ (main stem, branch orders, and leaves) for calculation of above-ground biomass production and allometric comparisons among aspen clones, species, and treatments. Additionally, we will collect fine root samples for DNA fingerprinting to quantify biomass production of individual aspen clones. This work will produce a thorough characterization of above-ground tree and stand growth and allocation above ground, and, in conjunction with the below ground harvest, total tree and stand biomass production, allocation, and allometry.

  6. Technical Review Report for the Safety Analysis Report for Packaging Model 9977 S-SARP-G-00001 Revision 2

    SciTech Connect (OSTI)

    DiSabatino, A; Hafner, R; West, M

    2007-10-04

    This Technical Review Report (TRR) summarizes the review findings for the Safety Analysis Report for Packaging (SARP) for the Model 9977 B(M)F-96 shipping container. The content analyzed for this submittal is Content Envelope C.1, Heat Sources, in assemblies of Radioisotope Thermoelectric Generators or food-pack cans. The SARP under review, i.e., S-SARP-G-00001, Revision 2 (August 2007), was originally referred to as the General Purpose Fissile Material Package. The review presented in this TRR was performed using the methods outlined in Revision 3 of the Department of Energy's (DOE's) Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's, Regulatory Guide 7.9, i.e., Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9977 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. The Model 9977 Package design includes a single, 6-inch diameter, stainless steel pressure vessel containment system (i.e., the 6CV) that was designed and fabricated in accordance with Section III, Subsection NB, of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code. The earlier package designs, i.e., the Model 9965, 9966, 9967 and 9968 Packages, were originally designed and certified in the 1980s. In the 1990s, updated package designs that incorporated design features consistent with new safety requirements, based on International Atomic Energy Agency guidelines, were proposed. The updated package designs were the Model 9972, 9973, 9974 and 9975 Packages, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under

  7. Safety overview of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.J.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.

    1996-05-23

    The National Ignition Facility (NIF) is a proposed US Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium- tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low.

  8. Safety overview of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.; Yatabe, J.

    1996-12-31

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory - New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996, after independent review. This paper summarizes the safety issues associated with the construction and operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low. 9 refs., 2 figs., 2 tabs.

  9. Technology-gap analysis of CNG refueling systems. Final report, July 1990-September 1991

    SciTech Connect (OSTI)

    Webb, R.F.

    1991-09-01

    The report provides a review and analysis of existing and emerging Compressed Natural Gas (CNG) refueling technology aimed at defining opportunities for improvements and areas where technical solutions might be sought. Interpretation of technical areas is broad, including not only scientific and engineering studies, laboratory work and technology demonstration (the usual areas for GRI support), but also technology transfer, support to develop and simplify regulations and economic analysis of technology options. The CNG refueling system is analyzed at several levels from an initial overview of the CNG market, at the area, refueling site, major equipment and component levels. The information has been used to generate a portfolio of 24 tasks for consideration by GRI in development of its future R and D program in support of CNG. The Appendix contains detail, references, a glossary and a report on the GRI Refueling Workshop held in Chicago January 16, 1991 (workshop findings are included in the main report but are not segregated from other findings).

  10. An economic analysis of a monitored retrievable storage site for Tennessee. Final report and appendices

    SciTech Connect (OSTI)

    Fox, W.F.; Mayo, J.W.; Hansen, L.T.; Quindry, K.E.

    1985-12-17

    The United States Department of Energy is charged with the task of identifying potential sites for a Monitored Retrievable Storage (MRS) Facility and reporting the results of its analysis to Congress by January 1986. DOE chose three finalist sites from 11 sites DOE analysts evaluated earlier. All three are in Tennessee, including two in Oak Ridge and one in Trousdale/Smith Counties. This paper is a summary of research undertaken on the economic effects of establishing the MRS facility in Tennessee. All three locations were considered in the analysis, but on some occasions attention is focused on the site preferred by DOE. The research was undertaken by the Center for Business and Economic Research (CBER), College of Business Administration, the University of Tennessee, Knoxville, under contract with the Tennessee Department of Economic and Community Development.

  11. Final Technical Report: Advanced Measurement and Analysis of PV Derate Factors.

    SciTech Connect (OSTI)

    King, Bruce Hardison; Burton, Patrick D.; Hansen, Clifford; Jones, Christian Birk

    2015-12-01

    The Advanced Measurement and Analysis of PV Derate Factors project focuses on improving the accuracy and reducing the uncertainty of PV performance model predictions by addressing a common element of all PV performance models referred to as “derates”. Widespread use of “rules of thumb”, combined with significant uncertainty regarding appropriate values for these factors contribute to uncertainty in projected energy production.

  12. Microsoft Word - Analysis WIPP Samples Integrated Summary Report FINAL.docx

    Office of Environmental Management (EM)

    LTR-2015/203 Oak Ridge National Laboratory Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report Phillip F. Britt March 2015 Approved for public release. Distribution is unlimited. DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect. Website http://www.osti.gov/scitech/ Reports produced before January 1, 1996, may be purchased by members of the public from the following source: National

  13. Microsoft Word - NRAP-TRS-III-006-2016_Parameter Sensitivity Analysis_final.2016.docx

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

    Parameter Sensitivity Analysis with the Seismicity Simulation Program RSQSim 28 January 2016 Office of Fossil Energy NRAP-TRS-III-006-2016 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information,

  14. Benefits analysis for the production of fuels and chemicals using solar thermal energy. Final report

    SciTech Connect (OSTI)

    1982-05-01

    Numerous possibilities exist for using high temperature solar thermal energy in the production of various chemicals and fuels (Sun Fuels). Research and development activities have focused on the use of feedstocks such as coal and biomass to provide synthesis gas, hydrogen, and a variety of other end-products. A Decision Analysis technique geared to the analysis of Sun Fuels options was developed. Conventional scoring methods were combined with multi-attribute utility analysis in a new approach called the Multi-Attribute Preference Scoring (MAPS) system. MAPS calls for the designation of major categories of attributes which describe critical elements of concern for the processes being examined. The six major categories include: Process Demonstration; Full-Scale Process, Feedstock; End-Product Market; National/Social Considerations; and Economics. MAPS calls for each attribute to be weighted on a simple scale for all of the candidate processes. Next, a weight is assigned to each attribute, thus creating a multiplier to be used with each individual value to derive a comparative weighting. Last, each of the categories of attributes themselves are weighted, thus creating another multiplier, for use in developing an overall score. With sufficient information and industry input, each process can be ultimately compared using a single figure of merit. After careful examination of available information, it was decided that only six of the 20 candidate processes were adequately described to allow a complete MAPS analysis which would allow direct comparisons for illustrative purposes. These six processes include three synthesis gas processes, two hydrogen and one ammonia. The remaining fourteen processes were subjected to only a partial MAPS assessment.

  15. Ultrascale Visualization Climate Data Analysis Tools (UV-CDAT) Final Report

    SciTech Connect (OSTI)

    Williams, Dean N.

    2014-05-19

    A partnership across government, academic, and private sectors has created a novel system that enables climate researchers to solve current and emerging data analysis and visualization challenges. The Ultrascale Visualization Climate Data Analysis Tools (UV-CDAT) software project utilizes the Python application programming interface (API) combined with C/C++/Fortran implementations for performance-critical software that offers the best compromise between "scalability" and “ease-of-use.” The UV-CDAT system is highly extensible and customizable for high-performance interactive and batch visualization and analysis for climate science and other disciplines of geosciences. For complex, climate data-intensive computing, UV-CDAT’s inclusive framework supports Message Passing Interface (MPI) parallelism as well as taskfarming and other forms of parallelism. More specifically, the UV-CDAT framework supports the execution of Python scripts running in parallel using the MPI executable commands and leverages Department of Energy (DOE)-funded general-purpose, scalable parallel visualization tools such as ParaView and VisIt. This is the first system to be successfully designed in this way and with these features. The climate community leverages these tools and others, in support of a parallel client-server paradigm, allowing extreme-scale, server-side computing for maximum possible speed-up.

  16. Safety Analysis Report for the use of hazardous production materials in photovoltaic applications at the National Renewable Energy Laboratory

    SciTech Connect (OSTI)

    Crandall, R.S.; Nelson, B.P.; Moskowitz, P.D.; Fthenakis, V.M.

    1992-07-01

    To ensure the continued safety of SERI`s employees, the community, and the environment, NREL commissioned an internal audit of its photovoltaic operations that used hazardous production materials (HPMs). As a result of this audit, NREL management voluntarily suspended all operations using toxic and/or pyrophoric gases. This suspension affected seven laboratories and ten individual deposition systems. These activities are located in Building 16, which has a permitted occupancy of Group B, Division 2 (B-2). NREL management decided to do the following. (1) Exclude from this SAR all operations which conformed, or could easily be made to conform, to B-2 Occupancy requirements. (2) Include in this SAR all operations that could be made to conform to B-2 Occupancy requirements with special administrative and engineering controls. (3) Move all operations that could not practically be made to conform to B-2 Occupancy requirements to alternate locations. In addition to the layered set of administrative and engineering controls set forth in this SAR, a semiquantitative risk analysis was performed on 30 various accident scenarios. Twelve presented only routine risks, while 18 presented low risks. Considering the demonstrated safe operating history of NREL in general and these systems specifically, the nature of the risks identified, and the layered set of administrative and engineering controls, it is clear that this facility falls within the DOE Low Hazard Class. Each operation can restart only after it has passed an Operational Readiness Review, comparing it to the requirements of this SAR, while subsequent safety inspections will ensure future compliance.

  17. Reduced order models for thermal analysis : final report : LDRD Project No. 137807.

    SciTech Connect (OSTI)

    Hogan, Roy E., Jr.; Gartling, David K.

    2010-09-01

    This LDRD Senior's Council Project is focused on the development, implementation and evaluation of Reduced Order Models (ROM) for application in the thermal analysis of complex engineering problems. Two basic approaches to developing a ROM for combined thermal conduction and enclosure radiation problems are considered. As a prerequisite to a ROM a fully coupled solution method for conduction/radiation models is required; a parallel implementation is explored for this class of problems. High-fidelity models of large, complex systems are now used routinely to verify design and performance. However, there are applications where the high-fidelity model is too large to be used repetitively in a design mode. One such application is the design of a control system that oversees the functioning of the complex, high-fidelity model. Examples include control systems for manufacturing processes such as brazing and annealing furnaces as well as control systems for the thermal management of optical systems. A reduced order model (ROM) seeks to reduce the number of degrees of freedom needed to represent the overall behavior of the large system without a significant loss in accuracy. The reduction in the number of degrees of freedom of the ROM leads to immediate increases in computational efficiency and allows many design parameters and perturbations to be quickly and effectively evaluated. Reduced order models are routinely used in solid mechanics where techniques such as modal analysis have reached a high state of refinement. Similar techniques have recently been applied in standard thermal conduction problems e.g. though the general use of ROM for heat transfer is not yet widespread. One major difficulty with the development of ROM for general thermal analysis is the need to include the very nonlinear effects of enclosure radiation in many applications. Many ROM methods have considered only linear or mildly nonlinear problems. In the present study a reduced order model is

  18. Switchgrass Biofuel Research: Carbon Sequestration and Life Cycle Analysis; Final Report

    SciTech Connect (OSTI)

    Liska, Adam J; Suyker, Andrew E; Arkebauer, Timothy J; Pelton, Matthew; Fang, Xiao Xue

    2013-12-20

    Soil emissions have been inadequately characterized in life cycle assessment of biofuels (see section 3.2.3). This project measures the net differences in field‐level greenhouse gas emissions (CO2, N2O, and CH4) due to corn residue removal for cellulosic ethanol production. Gas measurements are then incorporated into life cycle assessment of the final biofuel product to determine whether it is in compliance with federal greenhouse gas emissions standards for biofuels (Renewable Fuel Standard 2, RFS2). The field measurements have been conducted over three years on two, quarter‐section, production‐scale, irrigated corn fields (both roughly 50 hectares, as this size of field is necessary for reproducible eddy covariance flux measurements of CO2; chamber measurements are used to determine N2O and CH4 emissions). Due to a large hail storm in 2010, estimates of the emission from residue could not be separated from the total CO2 flux in 2011. This led us to develop soil organic carbon (SOC) modeling techniques to estimate changes in CO2 emissions from residue removal. Modeling has predicted emissions of CO2 from oxidation of SOC that are consistent (<12%) with 9 years of CO2 flux measurements at the two production field sites, and modeling is also consistent with other field measurements (Liska et al., submitted). The model was then used to estimate the average change in SOC and CO2 emissions from nine years of simulated residue removal (6 Mg biomass per hectare per year) at the sites; a loss of 0.43 Mg C ha‐1 yr‐1 resulted. The model was then used to estimate SOC changes over 10 years across Nebraska using supercomputing, based on 61 million, 30 x 30 meter, grid cells to account for regional variability in initial SOC, crop yield, and temperature; an average loss of 0.47 Mg C ha‐1 yr‐1 resulted. When these CO2 emissions are included in simple life cycle assessment calculations, emissions from cellulosic ethanol from crop residue are above mandated levels of

  19. Lift truck safety review

    SciTech Connect (OSTI)

    Cadwallader, L.C.

    1997-03-01

    This report presents safety information about powered industrial trucks. The basic lift truck, the counterbalanced sit down rider truck, is the primary focus of the report. Lift truck engineering is briefly described, then a hazard analysis is performed on the lift truck. Case histories and accident statistics are also given. Rules and regulations about lift trucks, such as the US Occupational Safety an Health Administration laws and the Underwriter`s Laboratories standards, are discussed. Safety issues with lift trucks are reviewed, and lift truck safety and reliability are discussed. Some quantitative reliability values are given.

  20. Final Environmental Impact Statement

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

    EIS - 0097-F c f ` f= �c J m s= Final Environmental Impact Statement BONNEVILLE POWER ADMINISTRATION TRANSMISSION FACILITIES VEGETATION MANAGEMENT PROGRAM U.S. Department of Energy August 1983 Appendices albLbfp= J MMVTJc= Responsible Official: tfiif^j= ^K= s^rde^k= Assistant Secretary for Environmental Protection. Safety, and Emergency Prepafedness Final Environmental Impact Statement BONNEVILLE POWER ADM IN ISTRA TION TRANSMISSION FACILITIES VEGETATION MANAGEMENT PROGRAM U.S. Department of

  1. The role of SASSYS-1 in LMR (Liquid Metal Reactor) safety analysis

    SciTech Connect (OSTI)

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs.

  2. FreedomCAR and vehicle technologies heavy vehicle program FY 2006. Benefits analysis : methodology and results - final report.

    SciTech Connect (OSTI)

    Singh, M.; Energy Systems; TA Engineering, Inc.

    2006-01-31

    This report describes the approach to estimating benefits and the analysis results for the Heavy Vehicle Technologies activities of the Freedom Car and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identification of technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in the activities planned for FY 06. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. These benefits estimates, along with market penetrations and other results, are then modeled as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY06 Budget Request.

  3. Freedom car and vehicle technologies heavy vehicle program : FY 2007 benefits analysis, methodology and results -- final report.

    SciTech Connect (OSTI)

    SIngh, M.; Energy Systems; TA Engineering

    2008-02-29

    This report describes the approach to estimating the benefits and analysis results for the Heavy Vehicle Technologies activities of the FreedomCar and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identifying technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in subsequent activities. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. The market penetrations are used as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY07 Budget Request. The energy savings models are utilized by the FCVT program for internal project management purposes.

  4. Energy engineering analysis program, Fort Polk, Louisiana. Pre-final executive summary, increment `f`

    SciTech Connect (OSTI)

    1987-11-01

    Executive Order 12003, dated 19 July 1977, initiated the U.S. Army`s energy conservation effort. Specifically, the Executive Order led to the development of the Army Facilities Energy Plan which directs Army Staff and Major Army Commands to develop detailed implementation plans for energy conservation. As a result of these directives, the Fort Worth District of the U.S. Army Corps of Engineers contracted for an Energy Engineering Analysis Program (EEAP) at Fort Polk, Louisiana. The EEAP included Increments `A`, `B`, `E`, and `O`.

  5. Analysis of Chinook Salmon in the Columbia River from an Ecosystem Perspective. Final Report.

    SciTech Connect (OSTI)

    Lichatowich, James A.; Mobrand, Lars E.

    1995-01-01

    Ecosystem Diagnosis and Treatment (EDT) methodology was applied to the analysis of chinook salmon in the mid-Columbia subbasins which flow through the steppe and steppe-shrub vegetation zones. The EDT examines historical changes in life history diversity related to changes in habitat. The emphasis on life history, habitat and historical context is consistent with and ecosystem perspective. This study is based on the working hypothesis that the decline in chinook salmon was at least in part due to a loss of biodiversity defined as the intrapopulation life history diversity. The mid Columbia subbasins included in the study are the Deschutes, John Day, Umatilla, Tucannon and Yakima.

  6. Final Report: Sublinear Algorithms for In-situ and In-transit Data Analysis at Exascale.

    SciTech Connect (OSTI)

    Bennett, Janine Camille; Pinar, Ali; Seshadhri, C.; Thompson, David; Salloum, Maher; Bhagatwala, Ankit; Chen, Jacqueline H.

    2015-09-01

    Post-Moore's law scaling is creating a disruptive shift in simulation workflows, as saving the entirety of raw data to persistent storage becomes expensive. We are moving away from a post-process centric data analysis paradigm towards a concurrent analysis framework, in which raw simulation data is processed as it is computed. Algorithms must adapt to machines with extreme concurrency, low communication bandwidth, and high memory latency, while operating within the time constraints prescribed by the simulation. Furthermore, in- put parameters are often data dependent and cannot always be prescribed. The study of sublinear algorithms is a recent development in theoretical computer science and discrete mathematics that has significant potential to provide solutions for these challenges. The approaches of sublinear algorithms address the fundamental mathematical problem of understanding global features of a data set using limited resources. These theoretical ideas align with practical challenges of in-situ and in-transit computation where vast amounts of data must be processed under severe communication and memory constraints. This report details key advancements made in applying sublinear algorithms in-situ to identify features of interest and to enable adaptive workflows over the course of a three year LDRD. Prior to this LDRD, there was no precedent in applying sublinear techniques to large-scale, physics based simulations. This project has definitively demonstrated their efficacy at mitigating high performance computing challenges and highlighted the rich potential for follow-on re- search opportunities in this space.

  7. Analysis of the premitting processes associated with exploration of Federal OCS leases. Final report

    SciTech Connect (OSTI)

    Not Available

    1980-11-01

    Under contract to the Office of Leasing Policy Development (LPDO), Jack Faucett Associates is currently undertaking the description and analysis of the Outer Continental Shelf (OCS) regulatory process to determine the nature of time delays that affect OCS production of oil and gas. This report represents the results of the first phase of research under this contract, the description and analysis of regulatory activity associated with exploration activities on the Federal OCS. Volume 1 contains the following three sections: (1) study results; (2) Federal regulatory activities during exploration of Federal OCS leases which involved the US Geological Survey, Environmental Protection Agency, US Coast Guard, Corps of Engineers, and National Ocean and Atmospheric Administration; and (3) state regulatory activities during exploration of Federal OCS leases of Alaska, California, Louisiana, Massachusetts, New Jersey, North Carolina, and Texas. Volume II contains appendices of US Geological Survey, Environmental Protection Agency, Coast Guard, Corps of Engineers, the Coastal Zone Management Act, and Alaska. The major causes of delay in the regulatory process governing exploration was summarized in four broad categories: (1) the long and tedious process associated with the Environmental Protection Agency's implementation of the National Pollutant Discharge Elimination System Permit; (2) the lack of mandated time periods for the completion of individual activities in the permitting process; (3) the lack of overall coordination of OCS exploratory regulation; and (4) the inexperience of states, the Federal government and industry relating to the appropriate level of regulation for first-time lease sale areas.

  8. Tools for Accurate and Efficient Analysis of Complex Evolutionary Mechanisms in Microbial Genomes. Final Report

    SciTech Connect (OSTI)

    Nakhleh, Luay

    2014-03-12

    I proposed to develop computationally efficient tools for accurate detection and reconstruction of microbes' complex evolutionary mechanisms, thus enabling rapid and accurate annotation, analysis and understanding of their genomes. To achieve this goal, I proposed to address three aspects. (1) Mathematical modeling. A major challenge facing the accurate detection of HGT is that of distinguishing between these two events on the one hand and other events that have similar "effects." I proposed to develop a novel mathematical approach for distinguishing among these events. Further, I proposed to develop a set of novel optimization criteria for the evolutionary analysis of microbial genomes in the presence of these complex evolutionary events. (2) Algorithm design. In this aspect of the project, I proposed to develop an array of e cient and accurate algorithms for analyzing microbial genomes based on the formulated optimization criteria. Further, I proposed to test the viability of the criteria and the accuracy of the algorithms in an experimental setting using both synthetic as well as biological data. (3) Software development. I proposed the nal outcome to be a suite of software tools which implements the mathematical models as well as the algorithms developed.

  9. Economic analysis of wind-powered farmhouse and farm building heating systems. Final report

    SciTech Connect (OSTI)

    Stafford, R.W.; Greeb, F.J.; Smith, M.F.; Des Chenes, C.; Weaver, N.L.

    1981-01-01

    The study evaluated the break-even values of wind energy for selected farmhouses and farm buildings focusing on the effects of thermal storage on the use of WECS production and value. Farmhouse structural models include three types derived from a national survey - an older, a more modern, and a passive solar structure. The eight farm building applications that were analyzed include: poultry-layers, poultry-brooding/layers, poultry-broilers, poultry-turkeys, swine-farrowing, swine-growing/finishing, dairy, and lambing. These farm buildings represent the spectrum of animal types, heating energy use, and major contributions to national agricultural economic values. All energy analyses were based on hour-by-hour computations which allowed for growth of animals, sensible and latent heat production, and ventilation requirements. Hourly or three-hourly weather data obtained from the National Climatic Center was used for the nine chosen analysis sites, located throughout the United States and corresponding to regional agricultural production centers.

  10. Methodology and application of surrogate plant PRA analysis to the Rancho Seco Power Plant: Final report

    SciTech Connect (OSTI)

    Gore, B.F.; Huenefeld, J.C.

    1987-07-01

    This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein.

  11. Final Meeting Summary Page 1

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

    June 18, 2015 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY, AND ENVIRONMENTAL PROTECTION COMMITTEE June 18, 2015 Richland, WA Topics in this Meeting Summary Opening ......................................................................................................................................................... 1 Tank Farm Vapors Advice Response ............................................................................................................ 1 Effectiveness of

  12. Technical Review Report for the Model 9975-96 Package Safety Analysis Report for Packaging (S-SARP-G-00003, Revision 0, January 2008)

    SciTech Connect (OSTI)

    West, M

    2009-05-22

    This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the U.S. Department of Energy (DOE), on the Safety Analysis Report for Packaging, Model 9975, Revision 0, dated January 2008 (S-SARP-G-00003, the SARP). The review includes an evaluation of the SARP, with respect to the requirements specified in 10 CFR 71, and in International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9975-96 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The safety analysis of the Model 9975-85 Packaging is documented in the Safety Analysis Report for Packaging, Model 9975, B(M)F-85, Revision 0, dated December 2003. The Model 9975-85 Package is certified by DOE Certificate of Compliance (CoC) package identification number, USA/9975/B(M)F-85, for the transportation of Type B quantities of uranium metal/oxide, {sup 238}Pu heat sources, plutonium/uranium metals, plutonium/uranium oxides, plutonium composites, plutonium/tantalum composites, {sup 238}Pu oxide/beryllium metal.

  13. Final Report Systems Level Analysis of the Function and Adaptive Responses of Methanogenic Consortia

    SciTech Connect (OSTI)

    Lovley, Derek R.

    2015-03-09

    The purpose of this research was to determine whether the syntrophic microbial associations that are central to the functioning of methane-producing terrestrial wetlands can be predictively modeled with coupled multi-species genome-scale metabolic models. Such models are important because methane is an important greenhouse gas and there is a need to predictively model how the methane-producing microbial communities will respond to environmental perturbations, such as global climate change. The research discovered that the most prodigious methane-producing microorganisms on earth participate in a previously unrecognized form of energy exchange. The methane-producers Methanosaeta and Methanosarcina forge biological electrical connections with other microbes in order to obtain electrons to reduce carbon dioxide to methane. This direct interspecies electron transfer (DIET) was demonstrated in complex microbial communities as well as in defined co-cultures. For example, metatranscriptomic analysis of gene expression in both natural communities and defined co-cultures demonstrated that Methanosaeta species highly expressed genes for the enzymes for the reduction of carbon dioxide to methane. Furthermore, Methanosaetas electron-donating partners highly expressed genes for the biological electrical connections known as microbial nanowires. A series of studies involving transcriptomics, genome resequencing, and analysis of the metabolism of a series of strains with targeted gene deletions, further elucidated the mechanisms and energetics of DIET in methane-producing co-cultures, as well as in a co-culture of Geobacter metallireducens and Geobacter sulfurreducens, which provided a system for studying DIET with two genetically tractable partners. Genome-scale modeling of DIET in the G. metallireducens/G. sulfurreducens co-culture suggested that DIET provides more energy to the electron-donating partner that electron exchange via interspecies hydrogen transfer, but that the

  14. 2007 Wholesale Power Rate Case Final Proposal : Risk Analysis Study Documentation.

    SciTech Connect (OSTI)

    United States. Bonneville Power Administration.

    2006-07-01

    The RiskMod Model is comprised of a set of risk simulation models, collectively referred to as RiskSim; a set of computer programs that manages data referred to as Data Management Procedures; and RevSim, a model that calculates net revenues. RiskMod interacts with the AURORA Model, the RAM2007, and the ToolKit Model during the process of performing the Risk Analysis Study. AURORA is the computer model being used to perform the Market Price Forecast Study (see Market Price Forecast Study, WP-07-FS-BPA-03); the RAM2007 is the computer model being used to calculate rates (see Wholesale Power Rate Development Study, WP-07-FS-BPA-05); and the ToolKit is the computer model being used to develop the risk mitigation package that achieves BPA's 92.6 percent TPP standard (see Section 3 in the Risk Analysis Study, WP-07-FS-BPA-04). Variations in monthly loads, resources, natural gas prices, forward market electricity prices, transmission expenses, and aluminum smelter benefit payments are simulated in RiskSim. Monthly spot market electricity prices for the simulated loads, resources, and natural gas prices are estimated by the AURORA Model. Data Management Procedures facilitate the format and movement of data that flow to and/or from RiskSim, AURORA, and RevSim. RevSim estimates net revenues using risk data from RiskSim, spot market electricity prices from AURORA, loads and resources data from the Load Resource Study, WP-07-FS-BPA-01, various revenues from the Revenue Forecast component of the Wholesale Power Rate Development Study, WP-07-FSBPA-05, and rates and expenses from the RAM2007. Annual average surplus energy revenues, purchased power expenses, and section 4(h)(10)(C) credits calculated by RevSim are used in the Revenue Forecast and the RAM2007. Heavy Load Hour (HLH) and Light Load Hour (LLH) surplus and deficit energy values from RevSim are used in the Transmission Expense Risk Model. Net revenues estimated for each simulation by RevSim are input into the Tool

  15. Facility Safety

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

    2005-12-22

    The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.

  16. Biological Safety

    Office of Energy Efficiency and Renewable Energy (EERE)

    The DOE's Biological Safety Program provides a forum for the exchange of best practices, lessons learned, and guidance in the area of biological safety. This content is supported by the Biosurety Executive Team. The Biosurety Executive Team is a DOE-chartered group. The DOE Office of Worker Safety and Health Policy provides administrative support for this group. The group identifies biological safety-related issues of concern to the DOE and pursues solutions to issues identified.

  17. Regulatory analysis on criteria for the release of patients administered radioactive material. Final report

    SciTech Connect (OSTI)

    Schneider, S.; McGuire, S.A.

    1997-02-01

    This regulatory analysis was developed to respond to three petitions for rulemaking to amend 10 CFR parts 20 and 35 regarding release of patients administered radioactive material. The petitions requested revision of these regulations to remove the ambiguity that existed between the 1-millisievert (0.1-rem) total effective dose equivalent (TEDE) public dose limit in Part 20, adopted in 1991, and the activity-based release limit in 10 CFR 35.75 that, in some instances, would permit release of individuals in excess of the current public dose limit. Three alternatives for resolution of the petitions were evaluated. Under Alternative 1, NRC would amend its patient release criteria in 10 CFR 35.75 to match the annual public dose limit in Part 20 of 1 millisievert (0.1 rem) TEDE. Alternative 2 would maintain the status quo of using the activity-based release criteria currently found in 10 CFR 35.75. Under Alternative 3, the NRC would revise the release criteria in 10 CFR 35.75 to specify a dose limit of 5 millisieverts (0.5 rem) TEDE.

  18. Probabilistic oil outflow analysis of alternative tanker designs. Addendum 1. Final report

    SciTech Connect (OSTI)

    Not Available

    1992-10-01

    This report is an addendum to Probabilistic Oil Outflow Analysis of Alternative Tanker Designs, U.S. Department of Transportation, U.S. Coast Guard, Report No. CG-D-14-92, July 1992. It covers the effects of design modifications to the Double Hull, Mid-Deck and Underpressure (MARPOL) designs for the 272,000 DWT size, and clarifies the evaluation made of the underpressure concept. The design modifications were made so that these ships would meet the requirements of Regulation 23 of the MARPOL 73/78 Annex for hypothetical outflow of oil. Results for two other designs, the Coulombi Egg and POLMIS, are also presented. A probabilistic approach has been adopted for these analyses, utilizing statistical data for tanker casualties developed for the International Maritime Organization. Both side and bottom damage were considered. Loss calculations were based on hydrostatic balance principles. Initial oil loss at impact and dynamic losses due to current and waves were computed based on model tests. Plots illustrating the cumulative oil outflow for each damage condition are presented. Three figures of merit were computed for each damage condition. These are: the probability of zero outflow, the mean outflow, and the extreme (1/10) outflow.

  19. Energy Engineering Analysis Program, Fort Polk, Louisiana. Final executive summary, increment `f`

    SciTech Connect (OSTI)

    1988-06-01

    Executive Order 12003, dated 19 July 1977, initiated the U.S. Army`s energy conservation effort. Specifically, the Executive Order led to the development of the Army Facilities Energy Plan which directs Army Staff and Major Army Commands to develop detailed implementation plans for energy conservation. As a result of these directives, the Fort Worth District of the U.S. Army Corps of Engineers contracted for an Energy Engineering Analysis Program (EEAP) at Fort Polk, Louisiana. The EEAP included Increments `A`, `B`, `E`, and `G`. To accomplish the intent of Increment `P`, namely, providing low cost/no cost energy savings recommendations in the form of specific, practical instructions for use by the Facility Engineer, the following general steps were taken: (1) Consider treasures identified in Detailed Scope of Work. (2) identify other potential Low Cost/No Cost energy Conservation Measures (ECM) through discussions with Fort Polk personnel and field surveys by Graham Associates engineers. (3) Review Increments `A`, `B`, and `G` for ECM`s within the Facility Engineer`s funding authority; $200,000 for alteration projects and $1,000,000 for maintenance and repair type work. (4) Evaluate ECM`s using relevant data for other Increments of the ESAP, and develop new data where appropriate.

  20. Major models and data sources for residential and commercial sector energy conservation analysis. Final report

    SciTech Connect (OSTI)

    Not Available

    1980-09-01

    Major models and data sources are reviewed that can be used for energy-conservation analysis in the residential and commercial sectors to provide an introduction to the information that can or is available to DOE in order to further its efforts in analyzing and quantifying their policy and program requirements. Models and data sources examined in the residential sector are: ORNL Residential Energy Model; BECOM; NEPOOL; MATH/CHRDS; NIECS; Energy Consumption Data Base: Household Sector; Patterns of Energy Use by Electrical Appliances Data Base; Annual Housing Survey; 1970 Census of Housing; AIA Research Corporation Data Base; RECS; Solar Market Development Model; and ORNL Buildings Energy Use Data Book. Models and data sources examined in the commercial sector are: ORNL Commercial Sector Model of Energy Demand; BECOM; NEPOOL; Energy Consumption Data Base: Commercial Sector; F.W. Dodge Data Base; NFIB Energy Report for Small Businesses; ADL Commercial Sector Energy Use Data Base; AIA Research Corporation Data Base; Nonresidential Buildings Surveys of Energy Consumption; General Electric Co: Commercial Sector Data Base; The BOMA Commercial Sector Data Base; The Tishman-Syska and Hennessy Data Base; The NEMA Commercial Sector Data Base; ORNL Buildings Energy Use Data Book; and Solar Market Development Model. Purpose; basis for model structure; policy variables and parameters; level of regional, sectoral, and fuels detail; outputs; input requirements; sources of data; computer accessibility and requirements; and a bibliography are provided for each model and data source.

  1. Lower Columbia River and Estuary Ecosystem Restoration Program Reference Site Study: 2011 Restoration Analysis - FINAL REPORT

    SciTech Connect (OSTI)

    Borde, Amy B.; Cullinan, Valerie I.; Diefenderfer, Heida L.; Thom, Ronald M.; Kaufmann, Ronald M.; Zimmerman, Shon A.; Sagar, Jina; Buenau, Kate E.; Corbett, C.

    2012-05-31

    The Reference Site (RS) study is part of the research, monitoring, and evaluation (RME) effort developed by the Action Agencies (Bonneville Power Administration [BPA], U.S. Army Corps of Engineers, Portland District [USACE], and U.S. Bureau of Reclamation) in response to Federal Columbia River Power System (FCRPS) Biological Opinions (BiOp). While the RS study was initiated in 2007, data have been collected at relatively undisturbed reference wetland sites in the LCRE by PNNL and collaborators since 2005. These data on habitat structural metrics were previously summarized to provide baseline characterization of 51 wetlands throughout the estuarine and tidal freshwater portions of the 235-km LCRE; however, further analysis of these data has been limited. Therefore, in 2011, we conducted additional analyses of existing field data previously collected for the Columbia Estuary Ecosystem Restoration Program (CEERP) - including data collected by PNNL and others - to help inform the multi-agency restoration planning and ecosystem management work underway in the LCRE.

  2. Pilot project for a commercial buildings Energy Analysis and Diagnostic Center (EADC) program. Final report

    SciTech Connect (OSTI)

    Capehart, B.L.

    1996-02-01

    Commercial energy use costs businesses around $70 billion annually. Many of these businesses are small and medium sized organizations that do not have the resources to help themselves, or to pay for professional engineering services to help reduce their energy costs and improve their economic competitiveness. Energy cost reduction actions with payback times of around two years could save the commercial sector 15--20%, or $10--$15 billion per year. This project was initially intended to evaluate the feasibility of performing commercial energy audits as an adjunct to the industrial audit program run by the US Department of Energy Industrial Office. This program is housed in 30 universities throughout the United States. Formerly known as Energy Analysis and Diagnostic Centers (EADC`s), the university programs are now called Industrial Assessment Centers (IAC`s) to reflect their expansion from energy use analyses to include waste and productivity analyses. The success of the EADC/IAC program in helping the manufacturing sector provides an excellent model for a similar program in the commercial buildings sector. This project has investigated using the EADC/IAC approach to performing energy audits for the commercial sector, and has determined that such an approach is feasible and cost effective.

  3. Health Safety and Environmental Protection Page 1

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

    Page 1 Final Meeting Summary March 10, 2011 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE March 10, 2011 Richland, WA Topics in this Meeting Summary Welcome and Introductions ............................................................................................................ 1 Beryllium ........................................................................................................................................ 1 Chemical

  4. SOLERAS - Solar Energy Water Desalination Project: Catalytic. System design final report. Volume 1. System requirement definition and system analysis

    SciTech Connect (OSTI)

    Not Available

    1986-01-01

    Catalytic Inc. was awarded a contract to conduct a preliminary system design and cost analysis for a brackish water desalination project to be located in Brownsville, Texas. System analyses and economic analyses were performed to define the baseline solar energy desalination system. The baseline system provides an average daily product water capacity of 6000 mT. The baseline system is optimal relative to technological risk, performance, and product water cost. Subsystems and their interfaces have been defined and product water cost projections made for commercial plants in a range of capacities. Science Applications, Inc. (SAI), subcontractor to Catalytic, had responsibility for the solar power system. This, the final report, summarizes the work performed under the Phase I effort.

  5. Facility Safety

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

    2005-12-22

    This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.

  6. Facility Safety

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

    2012-12-04

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.

  7. Facility Safety

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

    1995-10-13

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.

  8. Final Report

    SciTech Connect (OSTI)

    R Paul Drake

    2004-01-12

    OAK-B135 This is the final report from the project Hydrodynamics by High-Energy-Density Plasma Flow and Hydrodynamics and Radiation Hydrodynamics with Astrophysical Applications. This project supported a group at the University of Michigan in the invention, design, performance, and analysis of experiments using high-energy-density research facilities. The experiments explored compressible nonlinear hydrodynamics, in particular at decelerating interfaces, and the radiation hydrodynamics of strong shock waves. It has application to supernovae, astrophysical jets, shock-cloud interactions, and radiative shock waves.

  9. Genetic analysis of the regulation of TCH gene expression, Final Report

    SciTech Connect (OSTI)

    Braam, Janet

    2008-10-28

    investigation of CML24 function and regulation led to the finding that CML24 has a critical role in nitric oxide regulation. Distinct tilling mutant alleles demonstrated that CML24 can act as a switch in the response to day length perception. Because of potential redundancy with the related CML23 gene, CML23 T-DNA insertion mutants were identified and characterized. Together, CML23 and CML24 impact the autonomous regulatory pathway of the transition to flowering. Nitric oxide levels are elevated in cml23/cml24 double mutants. Therefore, CML23 and CML24 are potential calcium sensors regulate nitric oxide accumulation. In collaboration with Drs. McCann and Carpita, fourier transform infrared spectroscopy (FTIR) was used to assess, verify and classify wall architectural changes that occur as a result of single XTH insertion mutations. Thirty-four homozygous mutant lines of Arabidopsis representing 21 members of the xyloglucan endotransglucosylase/hydrolase gene family provided a set of mutants to characterize. Kohonen networks classified cell wall architectures of xth mutant lines and previously characterized cell wall mutants. The xth mutants were found to have chemical changes in their cell walls not detectable as phenotypic growth and development changes, consistent with the existence of feed-back loops that modify wall composition in response to a life-long deficiency of a cell wall enzyme. To gain insight into the potential physiological relevance of the distinct members of the XTH family, GUS reporter fusion genes were constructed, and plants expressing these transgenes were characterized to reveal spatial and temporal patterns of expression. In addition, Genevestigator sources were mined for comprehensive and comparative XTH expression regulation analysis. These data revealed that the Arabidopsis XTHs are likely expressed in every developmental stage from seed germination through flowering. All organs showed XTH::GUS expression and most, if not all, are found to express

  10. Model-based safety assessments

    SciTech Connect (OSTI)

    Carlson, D.D.; Jones, T.R.

    1998-04-01

    Sandia National Laboratories performs systems analysis of high risk, high consequence systems. In particular, Sandia is responsible for the engineering of nuclear weapons, exclusive of the explosive physics package. In meeting this responsibility, Sandia has developed fundamental approaches to safety and a process for evaluating safety based on modeling and simulation. These approaches provide confidence in the safety of our nuclear weapons. Similar concepts may be applied to improve the safety of other high consequence systems.

  11. Economic feasibility analysis of distributed electric power generation based upon the natural gas-fired fuel cell. Final report

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    The final report provides a summary of results of the Cost of Ownership Model and the circumstances under which a distributed fuel cell is economically viable. The analysis is based on a series of micro computer models estimate the capital and operations cost of a fuel cell central utility plant configuration. Using a survey of thermal and electrical demand profiles, the study defines a series of energy user classes. The energy user class demand requirements are entered into the central utility plant model to define the required size the fuel cell capacity and all supporting equipment. The central plant model includes provisions that enables the analyst to select optional plant features that are most appropriate to a fuel cell application, and that are cost effective. The model permits the choice of system features that would be suitable for a large condominium complex or a residential institution such as a hotel, boarding school or prison. Other applications are also practical; however, such applications have a higher relative demand for thermal energy, a characteristic that is well-suited to a fuel cell application with its free source of hot water or steam. The analysis combines the capital and operation from the preceding models into a Cost of Ownership Model to compute the plant capital and operating costs as a function of capacity and principal features and compares these estimates to the estimated operating cost of the same central plant configuration without a fuel cell.

  12. Final Scientific-Technical Report DOE-GISS-61768. Constraints on cloud feedback from analysis of arm observations and models

    SciTech Connect (OSTI)

    Del Genio, Anthony D.

    2015-04-15

    Final Scientific-Technical Report for research conducted under the Atmospheric Radiation Measurement Program from 1994-2010.

  13. Safety Analysis Report for Packaging (SARP): Model AL-M1 nuclear packaging (DOE C of C No. USA/9507/BLF)

    SciTech Connect (OSTI)

    Coleman, H.L.; Whitney, M.A.; Williams, M.A.; Alexander, B.M.; Shapiro, A.

    1987-11-24

    This Safety Analysis Report for Packaging (SARP) satisfies the request of the US Department of Energy for a formal safety analysis of the shipping container identified as USA/9507/BLF, also called AL-M1, configuration 5. This report makes available to all potential users the technical information and the limits pertinent to the construction and use of the shipping containers. It includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. A complete physical and technical description of the package is presented. The package consists of an inner container centered within an insulated steel drum. The configuration-5 package contains tritiated water held on sorbent material. There are two other AL-M1 packages, designated configurations 1 and 3. These use the same insulated outer drum, but licensing of these containers will not be addressed in this SARP. Design and development considerations, the tests and evaluations required to prove the ability of the container to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Monsanto Research Corporation-Mound experience in using the containers, and a copy of the DOE/OSD/ALO Certificate of Compliance are included.

  14. Nuclear Safety Analysis Reports

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

    1992-04-30

    Cancels DOE O 5481.1B; paragraphs 7b(3), 7e(3) & 8c of DOE O 5480.6; and 51, 7b(3), 7b(4), 7e(3), 8a & 8h of DOE O 5480.5.

  15. Analysis of NaOH releases for Hanford tank farms

    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: Caustic Spray Leak. The calculations needed to quantify the risk associated with this accident scenario are included within.

  16. Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

    Broader source: Energy.gov [DOE]

    During the 580th meeting of the Advisory Committee on Reactor Safeguards (ACRS), February10-12, 2011, we reviewed the staff’s white paper, “A Comparison of Integrated Safety Analysisand...

  17. SU-E-T-248: Near Real-Time Analysis of Radiation Delivery and Imaging, Accuracy to Ensure Patient Safety

    SciTech Connect (OSTI)

    Wijesooriya, K; Seitter, K; Desai, V; Read, P; Larner, J

    2014-06-01

    Purpose: To develop and optimize an effective software method for comparing planned to delivered control point machine parameters for all VARIAN TrueBeam treatments so as to permit (1) assessment of a large patient pool throughout their treatment course to quantify treatment technique specific systematic and random uncertainty of observables, (2) quantify the site specific daily imaging shifts required for target alignment, and (3) define tolerance levels for mechanical parameters and imaging parameters based on statistical analysis data gathered, and the dosimetric impact of variations. Methods: Treatment and imaging log files were directly compared to plan parameters for Eclipse and Pinnacle planned treatments via 3D, IMRT, control point, RapidArc, and electrons. Each control point from all beams/arcs (7984) for all fractions (1940) of all patients treated over six months were analyzed. At each control point gantry angle, collimator angle, couch angle, jaw positions, MLC positions, MU were compared. Additionally per-treatment isocenter shifts were calculated. Results were analyzed as a whole in treatment type subsets: IMRT, 3D, RapidArc; and in treatment site subsets: brain, chest/mediastinum, esophagus, H and N, lung, pelvis, prostate. Results: Daily imaging isocenter shifts from initial external tattoo alignment were dependent on the treatment site with < 0.5 cm translational shifts for H and N, Brain, and lung SBRT, while pelvis, esophagus shifts were ?1 cm. Mechanical delivery parameters were within tolerance levels for all sub-beams. The largest variations were for RapidArc plans: gantry angle 0.110.12,collimator angle 0.000.00, jaw positions 0.480.26, MLC leaf positions 0.660.08, MU 0.140.34. Conclusion: Per-control point validation reveals deviations between planned and delivered parameters. If used in a near real-time error checking system, patient safety can be improved by equipping the treatment delivery system with additional forcing functions

  18. Safety Issues

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

    Site Safety Orientation April, 2015 Atmospheric Radiation Measurement Climate Research Facility/ North Slope of Alaska/Adjacent Arctic Ocean (ACRF/NSA/AAO) Site Safety Orientation Purpose This document provides an overview and summary of safety issues and safe work practices associated with operations at the Atmospheric Radiation Measurement Climate Research Facility/North Slope of Alaska/Adjacent Arctic Ocean (ACRF/NSA/AAO) Sites. It is intended for site visitors as well as routine site

  19. Explosives Safety

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

    2012-06-27

    The Standard provides the basic technical requirements for an explosives safety program necessary for operations involving explosives, explosives assemblies, pyrotechnics and propellants, and assemblies containing these materials.

  20. Safety Engineer

    Broader source: Energy.gov [DOE]

    A successful candidate in this position will ensure DOE Federal personnel and contractors develop effective safety programs and continuously evaluates those activities to ensure compliance with DOE...

  1. Explosives Safety

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

    ... Schnetzer, Kimball O. Merewether, and Roy E. Jorgenson; May 1993 (current version 06 August 2002) Life Safety Code: NFPA 101, National Fire Protection Association (current version) ...

  2. Safety equipment list

    SciTech Connect (OSTI)

    Lavender, J.C.; Roe, N.A.

    1995-04-13

    This analysis assigns preliminary safety class (SC) designations to major systems and structures associated with the Multi-Function Waste Tank Facility (MWTF) project in accordance with the Multi-Function QAPP, W-236A (Hall 1994). Also included are SC assignments for those subsystems and major components of the major systems discussed in this document (see Appendices A and B). Component safety classifications have been completed through inspection (using engineering judgement) for simple systems, and through use of logic models (i.e., fault trees) for complicated systems. This analysis is intended to augment the SC systems list supplied in Chapter 9 of the MWTF Preliminary Safety Analysis Report (WHC 1994). Whereas WHC (1994) only addresses select systems, this analysis addresses the SC ramifications of all MWTF systems and structures as identified in the Title 1 design media. This document provides additional analyses of the system safety classifications assignments, and classifies systems not addressed in WHC (1994). This analysis specifically describes the safety functions) that must be performed by each MWTF system.

  3. Technical Review Report for the Model 9977 Safety Analysis Report for Packaging Addendum 1 Justification for DNDO Contents

    SciTech Connect (OSTI)

    West, M H

    2008-12-17

    The Model 9977 Package is currently certified for Content Envelope C.1, {sup 238}Pu Heat Sources, either in Radioisotope Thermoelectric Generator (RTG), or in Food-Pack Can configurations, under Certificate of Compliance (CoC) Certificate Number 9977 and Package Identification Number USA/9977/B(M)F-96 (DOE). Addendum 1, Justification for DNDO Contents,--the Submittal--supplements Revision 2 of the Safety Analysis Report for Packaging for the Model 9977 Package. The Submittal adds five new contents to the Model 9977 Package, Content Envelopes, AC.1 through AC.5. The Content Envelopes are neptunium metal, the beryllium-reflected plutonium ball (BeRP Ball), plutonium/uranium metal, plutonium/uranium metal with enhanced wt% {sup 240}Pu (to 50 wt%), and uranium metal. The last three Content Envelopes are stabilized to DOE-STD-3013. These Content Envelopes will be shipped to the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), where they will reside, and, hence, to off-site locations in support of the Department of Homeland Security (DHS) Domestic Nuclear Detection Office (DNDO). The new certificate will apply to a limited number of Model 9977 Packages. At the same time, the Submittal requests an extension of the periodic maintenance requirements from one (1) year to up to five (5) years using Radio-Frequency Identification (RFID) temperature-monitoring systems to measure the ambient storage temperature in order to ensure that the temperature of the Viton{reg_sign} O-rings for the 6-inch Containment Vessel (6CV) remain less than 200 F. The RFIDs have been developed by Argonne National Laboratory. An on-going surveillance program at the K-Area Materials Storage (KAMS) facility at the Savannah River Site, and an on-going examination of Viton{reg_sign} O-rings from mock Primary Containment Vessels (PCVs) at Savannah River National Laboratory (SRNL) provide the technical justification for the extension of the periodic maintenance interval. Where extended

  4. Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; Piyush Sabharwall; Rohit Upadhya

    2014-01-01

    For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial

  5. Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; Piyush Sabharwall; Rohit Upadhya

    2013-08-01

    For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial

  6. Refinement of procedures for analysis and construction of hydraulicked ash-slag dumps to improve their in-service safety

    SciTech Connect (OSTI)

    Frolov, A. N.

    2013-03-15

    Solutions are proposed for enhancement of the in-service safety of hydraulicked ash-slag dumps with consideration of their hydrothermal regime. An assessment is given for the minimum dimensions of the settling basins and top surface of ash-slag dumps.

  7. Final Optics

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

    final optics Final Optics Schematic layout of NIF's final optics assembly (FOA). The suite of optics for one beamline is on the right. The final optics assemblies (FOAs) are the last element of the main laser system and the first of the target area systems. Each FOA contains four integrated optics modules (IOMs) that incorporate beam conditioning, frequency conversion, focusing, diagnostic sampling, and debris shielding capabilities into a single compact assembly. These optics are shown in the

  8. Final technical report for Phenomic Analysis of Natural and Induced Variation in Brachypodium Distachyon DE-SC0001526

    SciTech Connect (OSTI)

    Vogel, John P.

    2014-12-17

    The goal of this project was to apply high-throughput, non-destructive phenotyping (phenomics) to collections of natural variants and induced mutants of the model grass Brachypodium distachyon and characterize a small subset of that material in detail. B. distachyon is well suited to this phenomic approach because its small size and rapid generation time allow researchers to grow many plants under carefully controlled conditions. In addition, the simple diploid genetics, high quality genome sequence and existence of numerous experimental tools available for B. distachyon allow us to rapidly identify genes affecting specific phenotypes. Our phenomic analysis revealed great diversity in biofuel-relevant traits like growth rate, biomass and photosynthetic rate. This clearly demonstrated the feasibility of applying a phenomic approach to the model grass B. distachyon. We also demonstrated the utility of B. distachyon for studying mature root system, something that is virtually impossible to do with biomass crops. We showed tremendous natural variation in root architecture that can potentially be used to design crops with superior nutrient and water harvesting capability. Finally, we demonstrated the speed with which we can link specific genes to specific phenotypes by studying two mutants in detail. Importantly, in both cases, the specific biological lessons learned were grass-specific and could not have been learned from a dicot model system. Furthermore, one of the genes affects cell wall integrity and thus may be a useful target in the context of biomass crop improvement. Ultimately, all this information can be used to accelerate the creation of improved biomass crops.

  9. Facility Safety

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

    2002-05-20

    To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.

  10. Transportation Safety

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

    Safety - 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 Energy Defense Waste Management Programs Advanced Nuclear Energy

  11. Analysis of the Relationship Between Vehicle Weight/Size and Safety, and Implications for Federal Fuel Economy Regulation

    SciTech Connect (OSTI)

    Wenzel, Thomas P.

    2010-03-02

    This report analyzes the relationship between vehicle weight, size (wheelbase, track width, and their product, footprint), and safety, for individual vehicle makes and models. Vehicle weight and footprint are correlated with a correlation coefficient (R{sup 2}) of about 0.62. The relationship is stronger for cars (0.69) than for light trucks (0.42); light trucks include minivans, fullsize vans, truck-based SUVs, crossover SUVs, and pickup trucks. The correlation between wheelbase and track width, the components of footprint, is about 0.61 for all light vehicles, 0.62 for cars and 0.48 for light trucks. However, the footprint data used in this analysis does not vary for different versions of the same vehicle model, as curb weight does; the analysis could be improved with more precise data on footprint for different versions of the same vehicle model. Although US fatality risk to drivers (driver fatalities per million registered vehicles) decreases as vehicle footprint increases, there is very little correlation either for all light vehicles (0.01), or cars (0.07) or trucks (0.11). The correlation between footprint and fatality risks cars impose on drivers of other vehicles is also very low (0.01); for trucks the correlation is higher (0.30), with risk to others increasing as truck footprint increases. Fatality risks reported here do not account for differences in annual miles driven, driver age or gender, or crash location by vehicle type or model. It is difficult to account for these factors using data on national fatal crashes because the number of vehicles registered to, for instance, young males in urban areas is not readily available by vehicle type or model. State data on all police-reported crashes can be used to estimate casualty risks that account for miles driven, driver age and gender, and crash location. The number of vehicles involved in a crash can act as a proxy of the number of miles a given vehicle type, or model, is driven per year, and is a

  12. The Development of Dynamic Human Reliability Analysis Simulations for Inclusion in Risk Informed Safety Margin Characterization Frameworks

    SciTech Connect (OSTI)

    Jeffrey C. Joe; Diego Mandelli; Ronald L. Boring; Curtis L. Smith; Rachel B. Shirley

    2015-07-01

    The United States Department of Energy is sponsoring the Light Water Reactor Sustainability program, which has the overall objective of supporting the near-term and the extended operation of commercial nuclear power plants. One key research and development (R&D) area in this program is the Risk-Informed Safety Margin Characterization pathway, which combines probabilistic risk simulation with thermohydraulic simulation codes to define and manage safety margins. The R&D efforts to date, however, have not included robust simulations of human operators, and how the reliability of human performance or lack thereof (i.e., human errors) can affect risk-margins and plant performance. This paper describes current and planned research efforts to address the absence of robust human reliability simulations and thereby increase the fidelity of simulated accident scenarios.

  13. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect (OSTI)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  14. Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    SciTech Connect (OSTI)

    Chang, T.Y.

    1991-09-01

    Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

  15. Safety is the First Priority

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

    is the First Priority The project team completed a number of overlapping safety methods to assure the safety of individuals operating and in proximity to the hydrogen fuel cell including: ● Failure Mode Effects Analysis (FMEA) which identifies potential failure points and devises ways to mitigate them (engineered and administrative/operational controls). ● Independent review and approval of the design by the Hydrogen Safety Panel and the US Coast Guard, and informational review by the

  16. Analysis of consequences of postulated solvent fires in Hanford site waste tanks

    SciTech Connect (OSTI)

    Cowley, W.L., Westinghouse Hanford

    1996-08-12

    This document contains the calculations that support the accident analyses for accidents involving organic solvents. This work was performed to support the Basis for Interim Operation (BIO) and the Final Safety Analysis Report (FSAR) for Tank Waste Remediation Systems (TWRS).

  17. Enterprise Assessments Targeted Review of the Safety Basis at...

    Energy Savers [EERE]

    River Site F-Area Central Laboratory Facility (referred to as the FH Laboratory) to evaluate changes in the Safety Analysis Report (SAR) and Technical Safety Requirements (TSRs). ...

  18. Updated laser safety&hazard analysis for the ARES laser system based on the 2007 ANSI Z136.1 standard.

    SciTech Connect (OSTI)

    Augustoni, Arnold L.

    2007-08-01

    A laser safety and hazard analysis was performed for the temperature stabilized Big Sky Laser Technology (BSLT) laser central to the ARES system based on the 2007 version of the American National Standards Institute's (ANSI) Standard Z136.1, for Safe Use of Lasers and the 2005 version of the ANSI Standard Z136.6, for Safe Use of Lasers Outdoors. The ARES laser system is a Van/Truck based mobile platform, which is used to perform laser interaction experiments and tests at various national test sites.

  19. 10 CFR part 851, Workers Safety and Health Program | Department...

    Office of Environmental Management (EM)

    CFR part 851, Workers Safety and Health Program 10 CFR part 851, Workers Safety and Health Program February 9, 2006 The Department published a final rule to implement the statutory ...

  20. Health, Safety and Environmental Protection Committee Page 1

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

    January 12, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE January 12, 2012 Richland, WA Topics in this Meeting Summary Welcome & Introductions ............................................................................................................................. 1 WTP Safety Culture Advice ......................................................................................................................... 1 Beryllium

  1. Health, Safety, and Environmental Protection Committee Page 1

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

    August 8, 2013 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY, AND ENVIRONMENTAL PROTECTION COMMITTEE August 8, 2013 Richland, WA Topics in this Meeting Summary Opening ......................................................................................................................................................... 1 Safety Culture Update ................................................................................................................................... 1

  2. Safety analysis report for the use of hazardous production materials in photovoltaic applications at the National Renewable Energy Laboratory. Volume 2, Appendices

    SciTech Connect (OSTI)

    Crandall, R.S.; Nelson, B.P.; Moskowitz, P.D.; Fthenakis, V.M.

    1992-07-01

    To ensure the continued safety of SERI`s employees, the community, and the environment, NREL commissioned an internal audit of its photovoltaic operations that used hazardous production materials (HPMS). As a result of this audit, NREL management voluntarily suspended all operations using toxic and/or pyrophoric gases. This suspension affected seven laboratories and ten individual deposition systems. These activities are located in Building 16, which has a permitted occupancy of Group B, Division 2 (B-2). NREL management decided to do the following. (1) Exclude from this SAR all operations which conformed, or could easily be made to conform, to B-2 Occupancy requirements. (2) Include in this SAR all operations that could be made to conform to B-2 Occupancy requirements with special administrative and engineering controls. (3) Move all operations that could not practically be made to conform to B-2 occupancy requirements to alternate locations. In addition to the layered set of administrative and engineering controls set forth in this SAR, a semiquantitative risk analysis was performed on 30 various accident scenarios. Twelve presented only routine risks, while 18 presented low risks. Considering the demonstrated safe operating history of NREL in general and these systems specifically, the nature of the risks identified, and the layered set of administrative and engineering controls, it is clear that this facility falls within the DOE Low Hazard Class. Each operation can restart only after it has passed an Operational Readiness Review, comparing it to the requirements of this SAR, while subsequent safety inspections will ensure future compliance. This document contains the appendices to the NREL safety analysis report.

  3. FINAL RELEASE

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

    RELEASE AWARDEE: ____________________________________________________ The work under Award No. DE-__________________________, dated ______________, between the United States of America (represented by the Department of Energy, National Energy Technology Laboratory, and the undersigned awardee, having been completed and finally accepted , and in consideration of Final Payment thereunder, the United States of America, its officers, agents and employees are hereby released from all liabilities,

  4. Technical Review Report for the Model 9978-96 Package Safety Analysis Report for Packaging (S-SARP-G-00002, Revision 1, March 2009)

    SciTech Connect (OSTI)

    West, M

    2009-03-06

    This Technical Review Report (TRR) documents the review, performed by Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the Department of Energy (DOE), on the 'Safety Analysis Report for Packaging (SARP), Model 9978 B(M)F-96', Revision 1, March 2009 (S-SARP-G-00002). The Model 9978 Package complies with 10 CFR 71, and with 'Regulations for the Safe Transport of Radioactive Material-1996 Edition (As Amended, 2000)-Safety Requirements', International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9978 Packaging is designed, analyzed, fabricated, and tested in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC). The review presented in this TRR was performed using the methods outlined in Revision 3 of the DOE's 'Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages'. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's Regulatory Guide 7.9, i.e., 'Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material'. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9978 Packaging is a single containment package, using a 5-inch containment vessel (5CV). It uses a nominal 35-gallon drum package design. In comparison, the Model 9977 Packaging uses a 6-inch containment vessel (6CV). The Model 9977 and Model 9978 Packagings were developed concurrently, and they were referred to as the General Purpose Fissile Material Package, Version 1 (GPFP). Both packagings use General Plastics FR-3716 polyurethane foam as insulation and as impact limiters. The 5CV is used as the Primary Containment Vessel (PCV) in the Model 9975-96 Packaging. The Model 9975-96 Packaging also has the 6CV as its Secondary Containment Vessel (SCV). In comparison, the Model 9975

  5. Establishing and Promoting a Culture of Safety in Chemistry Laboratory Research

    SciTech Connect (OSTI)

    Fryberger, Teresa

    2014-12-23

    Final technical/scientific report for the project, Establishing and Promoting a Culture of Safety in Chemistry Laboratory Research.

  6. Light-Weight Radioisotope Heater Unit final safety analysis report (LWRHU-FSAR): Volume 1: A. Introduction and executive summary: B. Reference Design Document (RDD)

    SciTech Connect (OSTI)

    Johnson, E.W.

    1988-10-01

    The orbiter and probe portions of the National Aeronautics and Space Administration (NASA) Galileo spacecraft contain components which require auxiliary heat during the mission. To meet these needs, the Department of Energy's (DOE's) Office of Special Applications (OSA) has sponsored the design, fabrication, and testing of a one-watt encapsulated plutonium dioxide-fueled thermal heater named the Light-Weight Radioisotope Heater Unit (LWRHU). This report, prepared by Monsanto Research Corporation (MRC), addresses the radiological risks which might be encountered by people both at the launch area and worldwide should postulated mission failures or malfunctions occur, resulting in the release of the LWRHUs to the environment. Included are data from the design, mission descriptions, postulated accidents with their consequences, test data, and the derived source terms and personnel exposures for the various events. 11 refs., 44 figs., 11 tabs.

  7. OSHA's Final Rule on Occupational Exposure to Respirable Crystalline...

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

    MB) More Documents & Publications Construction Safety Advisory Committee, Meeting Minutes - May 11, 2000 Federal Register Notice, CBDPP Final Rule Approaches to Safe Nanotechnology

  8. Final Report

    SciTech Connect (OSTI)

    Gurney, Kevin R

    2015-01-12

    This document constitutes the final report under DOE grant DE-FG-08ER64649. The organization of this document is as follows: first, I will review the original scope of the proposed research. Second, I will present the current draft of a paper nearing submission to Nature Climate Change on the initial results of this funded effort. Finally, I will present the last phase of the research under this grant which has supported a Ph.D. student. To that end, I will present the graduate student’s proposed research, a portion of which is completed and reflected in the paper nearing submission. This final work phase will be completed in the next 12 months. This final workphase will likely result in 1-2 additional publications and we consider the results (as exemplified by the current paper) high quality. The continuing results will acknowledge the funding provided by DOE grant DE-FG-08ER64649.

  9. Final Report

    SciTech Connect (OSTI)

    Biros, George

    2014-08-18

    This the final report for the project "Large-Scale Optimization for Bayesian Inference in Complex Systems," for the work in the group of the co-PI George Biros.

  10. Final Report

    SciTech Connect (OSTI)

    DeTar, Carleton

    2012-12-10

    This document constitutes the Final Report for award DE-FC02-06ER41446 as required by the Office of Science. It summarizes accomplishments and provides copies of scientific publications with significant contribution from this award.

  11. Preservation of FFTF Data Related to Passive Safety Testing

    SciTech Connect (OSTI)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

    2010-10-01

    experience to a large-size LMR and obtain data for validating design analysis computer codes, and 3) to develop and test passive safety enhancements that might be used for future LMRs. These tests were designed to provide data sufficient to allow separation of fuel temperature effects from structural temperature effects. The data developed through this testing program were used to verify the predictive capability of passive safety analysis methods as well as provide a data base for calibrating design tools such as the SASSYS/SAS4A codes. These tests were instrumental in improving understanding of reactivity feedback mechanisms in LMRs and demonstrating passive safety margins available in an LMR. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. This information may be of potential use for international exchanges with other LMR programs around the world. This information provides the basis for creating benchmarks for validating and testing large scale computer programs. All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. The test results information exists in several different formats depending upon the final stage of the test evaluation. Over 100 documents relevant to passive safety testing have been identified and are being recovered, scanned, and catalogued. Attempts to recover plant data tapes are also in progress. Documents related to passive safety testing are now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format compatible with a general search engine being developed by INL. The data from the FFTF passive safety tests provides experimental verification of structural reactivity effects that should be very useful to innovative designers seeking to optimize passive safety in the design of new LMRs.

  12. MINIMARS conceptual design: Final report

    SciTech Connect (OSTI)

    Lee, J.D.

    1986-09-01

    This volume contains the following sections: (1) fueling systems; (2) blanket; (3) alternative blanket concepts; (4) halo scraper/direct converter system study and final conceptual design; (5) heat-transport and power-conversion systems; (6) tritium systems; (7) minimars air detritiation system; (8) appropriate radiological safety design criteria; and (9) cost estimate. (MOW)

  13. Safety harness

    DOE Patents [OSTI]

    Gunter, Larry W.

    1993-01-01

    A safety harness to be worn by a worker, especially a worker wearing a plastic suit thereunder for protection in a radioactive or chemically hostile environment, which safety harness comprises a torso surrounding portion with at least one horizontal strap for adjustably securing the harness about the torso, two vertical shoulder straps with rings just forward of the of the peak of the shoulders for attaching a life-line and a pair of adjustable leg supporting straps releasibly attachable to the torso surrounding portion. In the event of a fall, the weight of the worker, when his fall is broken and he is suspended from the rings with his body angled slightly back and chest up, will be borne by the portion of the leg straps behind his buttocks rather than between his legs. Furthermore, the supporting straps do not restrict the air supplied through hoses into his suit when so suspended.

  14. 2012 Nuclear Safety Workshop Presentations

    Broader source: Energy.gov [DOE]

    Lists workshop presentations from: Wednesday, September 19 - Plenary Session Wednesday, September 19 - Beyond Design Basis Events Analysis and Response Breakout Session Wednesday, September 19 - Safety Culture Breakout Session Wednesday, September 19 - Risk Assessment and Management Breakout Session Thursday, September 20 - Beyond Design Basis Events Analysis and Response Breakout Session Thursday, September 20 - Safety Culture Breakout Session Thursday, September 20 - Risk Assessment and Management Breakout Session Thursday, September 20 - Plenary Session

  15. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  16. CSD Safety and Reliability Data

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

    CSD Safety & Reliability Data Forecourt CSD Workshop Argonne National Laboratory Kevin Harrison 20-Mar-2013 2 Approach - Composite Data Products CDPs DDPs Results Internal analysis completed quarterly Bundled data (operation & maintenance/safety) delivered to NREL quarterly Composite Data Products (CDPs) * Aggregated data across multiple systems, sites, and teams * Publish analysis results without revealing proprietary data every 6 months 2 Detailed Data Products (DDPs) * Individual data

  17. Health Safety and Environmental Protection Committee Page 1

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

    Page 1 Final Meeting Summary March 8, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE March 8, 2012 Richland, WA Topics in this Meeting Summary Welcome & Introductions ............................................................................................................................. 1 2012 Priority and Strategic Planning

  18. Health Safety and Environmental Protection Committee Page 1

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

    Committee Page 1 Final Meeting Summary May 10, 2011 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE May 10, 2011 Richland, WA Topics in this Meeting Summary Welcome and Introductions ............................................................................................................ 1 Committee Business........................................................................................................................ 8

  19. Safety valve

    DOE Patents [OSTI]

    Bergman, Ulf C.

    1984-01-01

    The safety valve contains a resilient gland to be held between a valve seat and a valve member and is secured to the valve member by a sleeve surrounding the end of the valve member adjacent to the valve seat. The sleeve is movable relative to the valve member through a limited axial distance and a gap exists between said valve member and said sleeve.

  20. Nuclear Safety Regulatory Framework

    Energy Savers [EERE]

    Authority and responsibility to regulate nuclear safety at DOE facilities 10 CFR 830 10 CFR 835 10 CFR 820 Regulatory Implementation Nuclear Safety Radiological Safety Procedural ...