National Library of Energy BETA

Sample records for documented safety analysis

  1. 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.

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

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

    5 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,...

  3. 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...

  4. 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)

  5. 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...

  6. 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.

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

    Office of Environmental Management (EM)

    should be addressed to: Office of Nuclear Safety (AU-30) Office of Environment, Health, ... The requirements for power, cooling water, and other external supplies to the ...

  8. 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.

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

    Broader source: Energy.gov [DOE]

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

  10. 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.

  11. WIPP Documents - Nuclear Safety

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

    Nuclear Safety DOE/WIPP-07-3372, Revision 4 WIPP Documented Safety Analysis Approved November 2013 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 4 WIPP Technical

  12. 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.

  13. 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.

  14. 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.

  15. 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, and keeping them as simple as possible while complying with industry standards and institutional requirements. No operations to be performed in the B695 Segment or building system are considered to be complex. No anticipated future change in the facility mission is expected to impact the extent of safety analysis documented in this DSA.

  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. 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).

  18. 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)

  19. Complete Experiment Safety Documentation

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

    Complete Experiment Safety Documentation Print User Safety Overview Upon receiving beam time: 1. Log in to ALSHub to complete an Experiment Safety Sheet (ESS). The ALS This e-mail address is being protected from spambots. You need JavaScript enabled to view it is available to support you through this process. Please contact This e-mail address is being protected from spambots. You need JavaScript enabled to view it at the email link or at (510) 486-7222 at if you have questions or need more

  20. Complete Experiment Safety Documentation

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

    Complete Experiment Safety Documentation Print User Safety Overview Upon receiving beam time: 1. Log in to ALSHub to complete an Experiment Safety Sheet (ESS). The ALS This e-mail address is being protected from spambots. You need JavaScript enabled to view it is available to support you through this process. Please contact This e-mail address is being protected from spambots. You need JavaScript enabled to view it at the email link or at (510) 486-7222 at if you have questions or need more

  1. Complete Experiment Safety Documentation

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

    Complete Experiment Safety Documentation Print User Safety Overview Upon receiving beam time: 1. Log in to ALSHub to complete an Experiment Safety Sheet (ESS). The ALS This e-mail address is being protected from spambots. You need JavaScript enabled to view it is available to support you through this process. Please contact This e-mail address is being protected from spambots. You need JavaScript enabled to view it at the email link or at (510) 486-7222 at if you have questions or need more

  2. Complete Experiment Safety Documentation

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

    Complete Experiment Safety Documentation Print User Safety Overview Upon receiving beam time: 1. Log in to ALSHub to complete an Experiment Safety Sheet (ESS). The ALS This e-mail address is being protected from spambots. You need JavaScript enabled to view it is available to support you through this process. Please contact This e-mail address is being protected from spambots. You need JavaScript enabled to view it at the email link or at (510) 486-7222 at if you have questions or need more

  3. 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.

  4. SSC Safety Review Document

    SciTech Connect (OSTI)

    Toohig, T.E. [ed.

    1988-11-01

    The safety strategy of the Superconducting Super Collider (SSC) Central Design Group (CDG) is to mitigate potential hazards to personnel, as far as possible, through appropriate measures in the design and engineering of the facility. The Safety Review Document identifies, on the basis of the Conceptual Design Report (CDR) and related studies, potential hazards inherent in the SSC project independent of its site. Mitigative measures in the design of facilities and in the structuring of laboratory operations are described for each of the hazards identified.

  5. 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.

  6. 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 and consideration at other DOE sites.

  7. Preparation Of Nonreactor Nuclear Facility Documented Safety...

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

    9-2014, Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis by Website Administrator This Department of Energy (DOE) Standard (STD), DOE-STD-3009-2014, describes...

  8. 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.

  9. 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.

  10. 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)

  11. Office of 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 Environmental Management (EM)

    Hanford Site Sludge Treatment Project Engineered Container Retrieval and Transfer System Preliminary Documented Safety Analysis, Revision 00 April 2015 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy Table of Contents Acronyms ................................................................................................................................. iii Executive Summary

  12. WIPP Documents - Quality Assurance and Safety

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

    Quality Assurance and Safety Quality Assurance Program Document DOECBFO-94-1012 Rev. 12 Effective date: 815...

  13. Nuclear Safety Enforcement Documents | Department of Energy

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

    Nuclear Safety Enforcement Documents Nuclear Safety Enforcement Documents Documents Available for Download April 28, 2016 Savannah River Nuclear Solutions, LLC, Consent Order NCO-2016-01 Nuclear Safety Enforcement Consent Order issued to Savannah River Nuclear Solutions, LLC relating to nuclear criticality safety infractions that occurred at the Savannah River Site February 19, 2016 Preliminary Notice of Violation, Los Alamos National Security, LLC Nuclear Safety Enforcement Preliminary Notice

  14. 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.

  15. Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

    SciTech Connect (OSTI)

    Dr. Michael A. Lehto; MAL

    2007-05-01

    A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFCs effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. 1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear safety requirements with waste requirements. MFCs efforts illustrate that utilizing the requirements of other disciplines, beyond nuclear safety, can provide an efficient process. Analyzing current processes to find better ways of meeting the requirements of multiple disciplines within a safety basis can lead to a more cost-effective, streamlined process. 2) Incorporating the DOT Type A drums into the MFC TSD was efficient because safety analysts utilized a transportation plan that provided analysis that could also be used for the change to the TSD addendum. In addition, because the plan they used had already been approved and was in use by the Idaho Cleanup Project (ICP) at the INL, justification for the change to the TSD was more compelling. MFC safety analysts proved that streamlining a process can be made more feasible by drawing from analysis that has already been completed.

  16. 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.

  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. 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. 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.

  1. 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

  2. Enterprise Assessments Operational Awareness Record for the Review of the WTP Low-Activity Waste Facility Preliminary Documented Safety Analysis Change Package for the Effluent Management Facility (OAR # EA-WTP-LAW-2016-01-25)

    Broader source: Energy.gov [DOE]

    Operational Awareness Record for the Review of the Waste Treatment and Immobilization Plant Low-Activity Waste Facility Preliminary Documented Safety Analysis Change Package for the Effluent Management Facility

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

    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.

  4. 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.

  5. 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...

  6. Criteria Document for B-plant Surveillance and Maintenance Phase Safety Basis Document

    SciTech Connect (OSTI)

    SCHWEHR, B.A.

    1999-08-31

    This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S&M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S&M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S&M Phase safety basis document.

  7. Nuclear Safety Enforcement Documents | Department of Energy

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

    Exposure at the Hanford Site June 14, 2005 Preliminary Notice of Violation, Safety and Ecology Corporation - EA-2005-03 Issued to Safety and Ecology Corporation related to a 10 CFR...

  8. Southern Great Plains Safety Orientation (Program Document) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect Program Document: Southern Great Plains Safety Orientation Citation Details In-Document Search Title: Southern Great Plains Safety Orientation Welcome to the Atmospheric Radiation Measurement (ARM) Climate Research Facility (ARM) Southern Great Plains (SGP) site. This U.S. Department of Energy (DOE) site is managed by Argonne National Laboratory (ANL). It is very important that all visitors comply with all DOE and ANL safety requirements, as well as those of the Occupational Safety

  9. Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Documents

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

    2014-12-19

    This Standard describes a framework and the criteria to be used for approval of (1) safety basis documents, as required by 10 Code of Federal Regulation (C.F.R.) 830, Nuclear Safety Management, and (2) safety design basis documents, as required by Department of Energy (DOE) Standard (STD)-1189-2008, Integration of Safety into the Design Process.

  10. Development of Onsite Transportation Safety Documents for Nevada Test Site

    SciTech Connect (OSTI)

    Frank Hand, Willard Thomas, Frank Sciacca, Manny Negrete, Susan Kelley

    2008-05-08

    Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if equivalent safety to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfers packaging and its contents.

  11. 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.

  12. 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.

  13. 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...

  14. 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...

  15. 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...

  16. Margin of Safety Definition and Examples Used in Safety Basis Documents and the USQ Process

    SciTech Connect (OSTI)

    Beaulieu, R. A.

    2013-10-03

    The Nuclear Safety Management final rule, 10 CFR 830, provides an undefined term, margin of safety (MOS). Safe harbors listed in 10 CFR 830, Table 2, such as DOE-STD-3009 use but do not define the term. This lack of definition has created the need for the definition. This paper provides a definition of MOS and documents examples of MOS as applied in a U.S. Department of Energy (DOE) approved safety basis for an existing nuclear facility. If we understand what MOS looks like regarding Technical Safety Requirements (TSR) parameters, then it helps us compare against other parameters that do not involve a MOS. This paper also documents parameters that are not MOS. These criteria could be used to determine if an MOS exists in safety basis documents. This paper helps DOE, including the National Nuclear Security Administration (NNSA) and its contractors responsible for the safety basis improve safety basis documents and the unreviewed safety question (USQ) process with respect to MOS.

  17. 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.

  18. FAQS Gap Analysis Qualification Card - Occupational Safety | Department

    Energy Savers [EERE]

    of Energy Occupational Safety FAQS Gap Analysis Qualification Card - Occupational Safety Functional Area Qualification Standard Gap Analysis Qualification Cards outline the differences between the last and latest version of the FAQ Standard. File Occupational Safety Gap Analysis Qualification Card More Documents & Publications FAQS Gap Analysis Qualification Card - Environmental Restoration FAQS Gap Analysis Qualification Card - Waste Management FAQS Gap Analysis Qualification Card -

  19. 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.

  20. Hanford Site Wide Transportation Safety Document [SEC 1 Thru 3

    SciTech Connect (OSTI)

    MCCALL, D L

    2002-06-01

    This safety evaluation report (SER) documents the basis for the US Department of Energy (DOE), Richland Operations Office (RL) to approve the Hanford Sitewide Transportation Safety Document (TSD) for onsite Transportation and Packaging (T&P) at Hanford. Hanford contractors, on behalf of DOE-RL, prepared and submitted the Hanford Sitewide Transportation Safety Document, DOE/RL-2001-0036, Revision 0, (DOE/RL 2001), dated October 4, 2001, which is referred to throughout this report as the TSD. In the context of the TSD, Hanford onsite shipments are the activities of moving hazardous materials, substances, and wastes between DOE facilities and over roadways where public access is controlled or restricted and includes intra-area and inter-area movements. The TSD sets forth requirements and standards for onsite shipment of radioactive and hazardous materials and wastes within the confines of the Hanford Site on roadways where public access is restricted by signs, barricades, fences, or other means including road closures and moving convoys controlled by Hanford Site security forces.

  1. Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

    SciTech Connect (OSTI)

    1995-01-01

    This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation.

  2. Cold Vacuum Drying facility design basis accident analysis documentation

    SciTech Connect (OSTI)

    CROWE, R.D.

    2000-08-08

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

  3. 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 evaluated and identified. This document supersedes the seismic classifications, assignments, and computations in ''Seismic Analysis for Preclosure Safety'' (BSC 2004a).

  4. Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Documents

    Energy Savers [EERE]

    SENSITIVE DOE-STD-1104-2009 May 2009 Superseding DOE-STD-1104-96 DOE STANDARD REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY BASIS AND SAFETY DESIGN BASIS DOCUMENTS U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-STD-1104-2009 ii Available on the Department of Energy Technical Standards web page at http://www.hss.energy.gov/nuclearsafety/ns/techstds/ DOE-STD-1104-2009 iii CONTENTS FOREWORD

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

    Office of Environmental Management (EM)

    2003 Technical Standards, Safety Analysis Toolbox Codes - November 2003 November 2003 Software Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes Safety...

  6. 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.

  7. Integration of Safety Culture Attributes into the EFCOG WP&C Program Guideline Document

    Broader source: Energy.gov [DOE]

    Slide Presentation by Steele Coddington, NSTec, Las Vegas and John McDonald, WRPS, Hanford. Integration of Safety Culture Attributes into EFCOG Work Planning and Control Guidance Document.

  8. 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 ...

  9. Review and Approval of Nuclear Facility Safety Basis Documents...

    Energy Savers [EERE]

    CHANGE NOTICE NO. 3 Date December 2005 DOE STANDARD REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY ... Neither a reviewer nor the preparer has veto power over ultimate resolution or ...

  10. Worker Safety and Health Enforcement Documents | Department of...

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

    ... LLC March 31, 2015 Consent Order, Fluor-B&W Portsmouth, LLC - March 12, 2015 Worker Safety and Health Enforcement Consent Order issued to Fluor-B&W Portsmouth, LLC for ...

  11. 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.

  12. Transient Safety Analysis of Fast

    Office of Scientific and Technical Information (OSTI)

    52 Transient Safety Analysis of Fast Spectrum tRu Burning LWRs with Internal Blankets Reactor Concepts Dr. Thomas Downar University of Michigan In collaboration with: Massachusetts Institute of Technology Argonne National Laboratory Thomas Sowinski, Federal POC Temitope Taiwo, Technical POC FINAL REPORT Project Title: Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets Covering Period: Final Date of Report: January 31, 2015 Recipient: University of Michigan 2355

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

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

    System Oversight Staffing Analysis - Example Safety System Oversight Staffing Analysis - Example Office spreadsheet icon SSO Alternate Staffing Analysis - Example Filled ...

  14. Hazardous Waste/Mixed Waste Treatment Building Safety Information Document (SID)

    SciTech Connect (OSTI)

    Fatell, L.B.; Woolsey, G.B.

    1993-04-15

    This Safety Information Document (SID) provides a description and analysis of operations for the Hazardous Waste/Mixed Waste Disposal Facility Treatment Building (the Treatment Building). The Treatment Building has been classified as a moderate hazard facility, and the level of analysis performed and the methodology used are based on that classification. Preliminary design of the Treatment Building has identified the need for two separate buildings for waste treatment processes. The term Treatment Building applies to all these facilities. The evaluation of safety for the Treatment Building is accomplished in part by the identification of hazards associated with the facility and the analysis of the facility`s response to postulated events involving those hazards. The events are analyzed in terms of the facility features that minimize the causes of such events, the quantitative determination of the consequences, and the ability of the facility to cope with each event should it occur. The SID presents the methodology, assumptions, and results of the systematic evaluation of hazards associated with operation of the Treatment Building. The SID also addresses the spectrum of postulated credible events, involving those hazards, that could occur. Facility features important to safety are identified and discussed in the SID. The SID identifies hazards and reports the analysis of the spectrum of credible postulated events that can result in the following consequences: Personnel exposure to radiation; Radioactive material release to the environment; Personnel exposure to hazardous chemicals; Hazardous chemical release to the environment; Events leading to an onsite/offsite fatality; and Significant damage to government property. The SID addresses the consequences to the onsite and offsite populations resulting from postulated credible events and the safety features in place to control and mitigate the consequences.

  15. Preparation plan, preliminary safety documentation, tank farm restoration and safe operations, Project W-314

    SciTech Connect (OSTI)

    Kidder, R.J.

    1994-10-20

    This preparation plan is developed to establish planning for the preliminary safety documentation for Project W-314, {open_quotes}Tank Farm Restoration and Safe Operations.{close_quotes}

  16. 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.

  17. Document

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

    or computer diskette) on request to the program contact person listed in this section. VIII. Other Information Electronic Access to This Document: You may view this document, ...

  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. Rapid Exploitation and Analysis of Documents

    SciTech Connect (OSTI)

    Buttler, D J; Andrzejewski, D; Stevens, K D; Anastasiu, D; Gao, B

    2011-11-28

    Analysts are overwhelmed with information. They have large archives of historical data, both structured and unstructured, and continuous streams of relevant messages and documents that they need to match to current tasks, digest, and incorporate into their analysis. The purpose of the READ project is to develop technologies to make it easier to catalog, classify, and locate relevant information. We approached this task from multiple angles. First, we tackle the issue of processing large quantities of information in reasonable time. Second, we provide mechanisms that allow users to customize their queries based on latent topics exposed from corpus statistics. Third, we assist users in organizing query results, adding localized expert structure over results. Forth, we use word sense disambiguation techniques to increase the precision of matching user generated keyword lists with terms and concepts in the corpus. Fifth, we enhance co-occurrence statistics with latent topic attribution, to aid entity relationship discovery. Finally we quantitatively analyze the quality of three popular latent modeling techniques to examine under which circumstances each is useful.

  20. 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.

  1. Document

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

    171 Federal Register / Vol. 75, No. 108 / Monday, June 7, 2010 / Notices an accessible format (e.g., braille, large print, audiotape, or computer diskette) on request to the contact person listed under FOR FURTHER INFORMATION CONTACT. Electronic Access to This Document You can view this document, as well as all other documents of this Department published in the Federal Register, in text or Adobe Portable Document Format (PDF) on the Internet at the following site: http://www.ed.gov/

  2. 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.

  3. Exploratory Nuclear Reactor Safety Analysis and Visualization via

    Office of Scientific and Technical Information (OSTI)

    Integrated Topological and Geometric Techniques (Technical Report) | SciTech Connect Technical Report: Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques Citation Details In-Document Search Title: Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming

  4. Implementation Guide for Use in Developing Documented Safety Analyses to Meet Subpart B of 10 CFR 830

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

    2001-10-24

    This Guide elaborates on the documented safety analysis (DSA) development process and the safe harbor provisions of the Appendix to10 CFR 830 Subpart B. Title 10 Code of Federal Regulations (CFR) Part 830, Subpart B, 'Safety Basis Requirements,' requires the contractor responsible for a Department of Energy (DOE) nuclear facility to analyze the facility, the work to be performed, and the associated hazards and to identify the conditions, safe boundaries, and hazard controls necessary to protect workers, the public, and the environment from adverse consequences. Canceled by DOE G 421.1-2A

  5. Test Protocol Document, Hydrogen Safety Sensor Testing; Phase I: Non-Flammable Mixtures

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

    Test Protocol Document, Hydrogen Safety Sensor Testing Phase I: Non-Flammable Mixtures R. Burgess, C. Blake, and C.E. Tracy Technical Report NREL/TP-560-42666 September 2008 NREL is operated by Midwest Research Institute ● Battelle Contract No. DE-AC36-99-GO10337 Test Protocol Document, Hydrogen Safety Sensor Testing Phase I: Non-Flammable Mixtures R. Burgess, C. Blake, and C.E. Tracy Prepared under Task No. H274.7110 Technical Report NREL/TP-560-42666 September 2008 National Renewable Energy

  6. Documents

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

    Summation version Stakeholder Summation version DOE Documents Environmental Management SSAB 15 Year Anniversary Paper (presented at the Waste Management Symposium, March 4, 2009)...

  7. Battery Lifetime Analysis and Simulation Tool (BLAST) Documentation

    Office of Scientific and Technical Information (OSTI)

    Battery Lifetime Analysis and Simulation Tool (BLAST) Documentation Neubauer, J. 25 ENERGY STORAGE BATTERY; LITHIUM-ION; STATIONARY ENERGY STORAGE; BLAST; BATTERY DEGRADATION;...

  8. Document

    National Nuclear Security Administration (NNSA)

    of the Draft EIS was also published in four local and regional newspapers (in English and Spanish) and over 200 copies of the document were distributed to federal, state...

  9. Document

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

    published in the Federal Register, in text or Adobe Portable Document Format (PDF) on the Internet at the following site: http:www.ed.gov newsfedregister. To use PDF you must...

  10. Document

    National Nuclear Security Administration (NNSA)

    ... is available on the Internet through the NEPA Web site at http:www.gc.energy.govnepa. ... internet Web site at http:www.ferc.gov under the link to Documents and Filings. ...

  11. Document

    National Nuclear Security Administration (NNSA)

    The Code of Federal Regulations is sold by the Superintendent of Documents. Prices of new books are listed in the first FEDERAL REGISTER issue of each week. Rules and Regulations ...

  12. Documents

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

    Documents Documents Sort by: Default | Name | Date (low-high) | Date (high-low) | Source | Category White Paper Template May 7, 2015 | Download File: HEP-White-Papers-Public2.pdf | pdf | 78 KB Download File: ASCRHEPReqRevWPtemplate.docx | docx | 136 KB HEP Exascale Requirements Review Case Study Template June 2, 2015 | Download File: CaseStudyTemplate-HEPrev.docx | docx | 28 KB HEP Exascale Requirements Review Case Study Template Last edited: 2016-04-29 11:34:52

  13. 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 ...

  14. 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.

  15. 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

  16. Document

    National Nuclear Security Administration (NNSA)

    1850 Federal Register / Vol. 75, No. 137 / Monday, July 19, 2010 / Notices 1 The NOI identified the title of the document as the Supplemental Environmental Impact Statement for Surplus Plutonium Disposition at the Savannah River Site. Issued in Washington, DC, on July 9, 2010. Cathy Zoi, Assistant Secretary, Energy Efficiency and Renewable Energy. [FR Doc. 2010-17514 Filed 7-16-10; 8:45 am] BILLING CODE 6450-01-P DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Project No.

  17. Document

    National Nuclear Security Administration (NNSA)

    60017 Federal Register / Vol. 76, No. 188 / Wednesday, September 28, 2011 / Notices 1 Enforcement of Statutes, Orders, Rules, and Regulations, 132 FERC ¶ 61,216 (2010). Persons who wish to comment only on the environmental review of this project should submit an original and two copies of their comments to the Secretary of the Commission. Environmental commentors will be placed on the Commission's environmental mailing list, will receive copies of the environmental documents, and will be

  18. Document

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

    ... reports, financial and product trend analysis, and other related records. Cut off annually. Destroy 3 years after cutoff. (N1-434-95-5) d. Product, Services Marketing, and ...

  19. Document

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

    at 202-564-7146 or http:www.epa.gov compliancenepa. ... Alternative Analysis, U.S. COE Section 404 Permit, ...28 2010, Contact: Allison Arnold, 512- 490-0057 ...

  20. Document:

    Office of Legacy Management (LM)

    DATA DECISION ANALYSIS: Project Shoal prepared by Greg Pohll, John Tracy, Frank Forsgren submitted to Nevada Operations Office U.S. Department of Energy January 1999 DOE/NV/11508-42 Publication No. 45166 This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors or their employees, makes any warranty, express or implied, or

  1. 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.

  2. 10 CFR 851 Worker Safety and Health Program (WSHP) Description Document

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

    Oak Ridge Operations P.O. Box 2001 Oak Ridge, Tennessee 37831- May 2 2, 20 0 7 Dr. Ronald D. Townsend, Director Oak Ridge Institute for Science and Education Mai! Stop 22 Post Office Box I 17 Oak Ridge, Tennessee 37831-0117 Dear Dr. Townsend: CONTRACT NO. DEAC05-06OR23100, OAK RIDGE INSTITUTE FOR SCIENCE AND EDUCATION (ORISE) 10 CFR 851 WORKER SAFETY AND HEALTH PROGRAM (WSHP) DESCRIPTION DOCUMENT The ORISE WSHP submitted to us on February 26, 2007, was evaluated to provide reasonable assurance

  3. Plant perspective on electronic document management systems for process safety management programs

    SciTech Connect (OSTI)

    Mannan, M.; Lee, W.S. [RMT/Jones and Neuse, Inc., Austin, TX (United States)

    1996-08-01

    As mandated by the Clean Air Act Amendments of 1990, the Occupational Safety and Health Administration (OSHA) promulgated on February 24, 1992 the Process Management (PSM) rule: 29 CFR 1910.119. Compliance with the PSM requirements is causing a significant impact on both financial resources and manpower for the covered facilities. A major portion of the efforts have been concentrated on developing and compiling a vast amount of data in various formats (i.e., autocad, word processing, and numeric databases). Because of the need to access and revise these data on a continuous basis, many facilities have started to implement computer-based document management systems to store, manage, and maintain these data. This paper provides a plant perspective of the necessary features and characteristics of a computer-based document management system. An electronic document management system can also form the basis for a plant workflow system. Once, all the information related to a plant`s PSM program are inputted into the system, it is possible to control the plant workflow using the management of change system and maintenance work requests. Other key issues that must be included in a computer-based system are redlining capabilities, execution of management of change authorization, and interfacing with existing computer programs (many plants are already using different kinds of software for a variety of objectives). This paper provides a roadmap for the development and implementation of an electronic document management system.

  4. 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}.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Document Reviews

    Broader source: Energy.gov [DOE]

    The Office of Document Reviews, within the Office of Health, Safety and Security information and document declassification efforts promote the release of information needed by an informed citizenry...

  10. Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)

    SciTech Connect (OSTI)

    1986-12-01

    With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.

  11. 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 ...

  12. Documents - Hanford Site

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

    Documents Documents Documents Documents Hanford Site Cleanup Completion Framework Tri-Party Agreement Freedom of Information and Privacy Act Hanford Site Budget Hanford Site Safety Standards DOE - ORP Contracts/Procurements DOE - RL Contracts/Procurements Integrated Waste Feed Delivery Plan Single-Shell Tank Evaluations Deep Vadose Zone 100-F RI/FS Sitewide Probabilistic Seismic Hazard Analysis Environmental Documents Email Email Page | Print Print Page |Text Increase Font Size Decrease Font

  13. 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.

  14. Preparation of Documented Safety Analysis for Interim Operations...

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

    ... public, or environment, including: (1) physical design, structural, and engineering ... such as meteorological data, air dispersion modelling, or natural phenomena hazards). ...

  15. 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.

  16. 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.

  17. 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.

  18. Criticality Safety | Department of Energy

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

    Services » Nuclear Safety » Criticality Safety Criticality Safety Nuclear Safety Basis The Nuclear Facility Safety Program establishes and maintains the DOE requirements for nuclear criticality safety. The DOE detailed requirements for criticality safety are contained in Section 4.3 of the DOE Order 420.1,Facility Safety. Criticality safety requirements are based on the documented safety analysis required by 10 CFR 830, Subpart B. Related Links 10 CFR 830, Nuclear Safety Management American

  19. Documents - Hanford Site

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

    Documents Documents Documents Hanford Site Cleanup Completion Framework Tri-Party Agreement Freedom of Information and Privacy Act Hanford Site Budget Hanford Site Safety Standards DOE - ORP Contracts/Procurements DOE - RL Contracts/Procurements Integrated Waste Feed Delivery Plan Single-Shell Tank Evaluations Deep Vadose Zone 100-F RI/FS Sitewide Probabilistic Seismic Hazard Analysis Environmental Documents Email Email Page | Print Print Page |Text Increase Font Size Decrease Font Size

  20. 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.

  1. 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...

  2. 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.

  3. Volume II - Accident and Operational Safety Analysis Handbook

    Energy Savers [EERE]

    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

  4. 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).

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

    Office of Scientific and Technical Information (OSTI)

    Blankets (Technical Report) | SciTech Connect 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 LWRs with Internal Blankets 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

  6. 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 include April 2007 through March 2008. The results from analyzing the deficiencies are presented in accordance with the two primary NTS reporting thresholds for WSH and nuclear safety noncompliances: (1) those related to certain events or conditions and (2) those that are management issues. In addition, WSH noncompliances were also analyzed to determine if any fell under the 'other significant condition' threshold. This report also identifies noncompliance topical areas that may have issues that do not meet the NTS reporting threshold but should remain under observation. These are placed on the 'watch list' for continued analysis.

  7. 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.

  8. GCtool for fuel cell systems design and analysis : user documentation.

    SciTech Connect (OSTI)

    Ahluwalia, R.K.; Geyer, H.K.

    1999-01-15

    GCtool is a comprehensive system design and analysis tool for fuel cell and other power systems. A user can analyze any configuration of component modules and flows under steady-state or dynamic conditions. Component models can be arbitrarily complex in modeling sophistication and new models can be added easily by the user. GCtool also treats arbitrary system constraints over part or all of the system, including the specification of nonlinear objective functions to be minimized subject to nonlinear, equality or inequality constraints. This document describes the essential features of the interpreted language and the window-based GCtool environment. The system components incorporated into GCtool include a gas flow mixer, splitier, heater, compressor, gas turbine, heat exchanger, pump, pipe, diffuser, nozzle, steam drum, feed water heater, combustor, chemical reactor, condenser, fuel cells (proton exchange membrane, solid oxide, phosphoric acid, and molten carbonate), shaft, generator, motor, and methanol steam reformer. Several examples of system analysis at various levels of complexity are presented. Also given are instructions for generating two- and three-dimensional plots of data and the details of interfacing new models to GCtool.

  9. Risk D&D Rapid Prototype: Scenario Documentation and Analysis Tool

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Seiple, Timothy E.

    2009-05-28

    Report describes process and methodology associated with a rapid prototype tool for integrating project risk analysis and health & safety risk analysis for decontamination and decommissioning projects.

  10. 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.

  11. 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 ...

  12. 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...

  13. 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.

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

    Energy Savers [EERE]

    Planning and Control, Final Report | Department of Energy 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 Department of

  15. Monthly Analysis of Electrical Safety Occurrences - January 2012 |

    Energy Savers [EERE]

    Department of Energy 2 Monthly Analysis of Electrical Safety Occurrences - January 2012 January 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 January 2012 and past information for information exchange and continual learning. PDF icon Monthly Analysis of Electrical Safety Occurrences - January 2012 More

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

    Energy Savers [EERE]

    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. PDF icon Monthly Analysis of Electrical Safety Occurrences - October 2012 More

  17. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations.

  18. Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

    SciTech Connect (OSTI)

    Smith, Paul Herrick; Teague, Jonathan Gayle

    2015-04-30

    Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

  19. 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.

  20. Monthly Analysis of Electrical Safety Occurrences – February 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.

  1. Monthly Analysis of Electrical Safety Occurrences – September 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.

  2. 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.

  3. 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.

  4. 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.

  5. Monthly Analysis of Electrical Safety Occurrences – September 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.

  6. 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.

  7. 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.

  8. Monthly Analysis of Electrical Safety Occurrences – July 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.

  9. 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.

  10. 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.

  11. 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.

  12. Monthly Analysis of Electrical Safety Occurrences – March 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 – 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Monthly Analysis of Electrical Safety Occurrences – November 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 – 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.

  20. Monthly Analysis of Electrical Safety Occurrences – July 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.

  1. 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.

  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 – 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Safety assessment document (SAD) for the Princeton Beta Experiment Modification (PBX-M)

    SciTech Connect (OSTI)

    Stencel, J.R.; Parsells, R.F.

    1988-04-01

    The Princeton Beta Experiment-Modification (PBX-M) is an experimental device of the tokamak type. A tokamak is characterized by a strong toroidal magnetic field composed of an externally driven component parallel to the torus centerline modified by the field produced by a transformer-driven current (OH) in the confined plasma. A second magnetic field parallel to the major toroidal axis is added to provide radial equilibrium for the plasma. As an advanced tokamak, PBX-M will have additional magnetic fields to reshape the plasma cross section from a circle into a kidney bean shape; it will also be equipped with 6MW or more of auxiliary heating power provided by four neutral beam injectors, with RF systems, and with an extensive set of diagnostics. Potential hazards associated with PBX-M, which are analyzed in this report, result from energy stored in the magnetic fields, high voltages necessary for the operation of some of the equipment and diagnostics, neutron radiation when the neutral beams are run with deuterium and x-rays, especially those emitted as a result of plasma-wall interaction. This report satisfies the requirements set forth in the PPPL Health and Safety Directives, specifically HSD-5003, and in DOE Order 5481.1B and its Chicago operations supplement (DOE86, DOE82).

  9. 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.

  10. Technical support document: Survey data used in revising the energy conservation requirements in the manufactured home construction and safety standards

    SciTech Connect (OSTI)

    Lee, A.D.; Conner, C.C.

    1992-02-01

    This report documents data used in the development of revised energy conservation standards for manufactured housing. The approach used in developing the proposed standard revision is a cost-benefit analysis in which the costs of energy conservation measures (ECM) are balanced against the benefits of energy savings. The analysis used to develop the recommendations for revision of the US Department of Housing and Urban Development (HUD) energy conservation standards for manufactured housing requires information on specific ECMs. This technical support document contains the data from two of the three surveys that were used primarily to characterize the available ECMs and their associated costs. The analyses of these data are provided in separate reports. 5 refs.

  11. ARM - Publications: Science Team Meeting Documents: Using EOF analysis to

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

    uncover inhomogeneities in data from ground-based aerosol monitoring devices Using EOF analysis to uncover inhomogeneities in data from ground-based aerosol monitoring devices Gianelli, Scott NASA/GISS Carlson, Barbara NASA/Goddard Institute for Space Studies Lacis, Andrew NASA/Goddard Institute for Space Studies Empirical Orthogonal Function (EOF) analysis is performed on ground-based shadowband and sun photometer data. The instruments examined for this study include the Rotating Shadowband

  12. Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and

    Energy Savers [EERE]

    Example Sheet) | Department of Energy Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and Example Sheet) Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and Example Sheet) This 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. Use caution when making worksheet modifications since changes to the

  13. Analysis and decision document in support of acquisition of steam supply for the Hanford 200 Area

    SciTech Connect (OSTI)

    Brown, D.R.; Daellenbach, K.K.; Hendrickson, P.L.; Kavanaugh, D.C.; Reilly, R.W.; Shankle, D.L.; Smith, S.A.; Weakley, S.A.; Williams, T.A. ); Grant, T.F. )

    1992-02-01

    The US Department of Energy (DOE) is now evaluating its facility requirements in support of its cleanup mission at Hanford. One of the early findings is that the 200-Area steam plants, constructed in 1943, will not meet future space heating and process needs. Because the 200 Area will serve as the primary area for waste treatment and long-term storage, a reliable steam supply is a critical element of Hanford operations. This Analysis and Decision Document (ADD) is a preliminary review of the steam supply options available to the DOE. The ADD contains a comprehensive evaluation of the two major acquisition options: line-term versus privatization. It addresses the life-cycle costs associated with each alternative, as well as factors such as contracting requirements and the impact of market, safety, security, and regulatory issues. Specifically, this ADD documents current and future steam requirements for the 200 Area, describes alternatives available to DOE for meeting these requirements, and compares the alternatives across a number of decision criteria, including life-cycle cost. DOE has currently limited the ADD evaluation alternatives to replacing central steam plants rather than expanding the study to include alternative heat sources, such as a distributed network of boilers or heat pumps. Thirteen project alternatives were analyzed in the ADD. One of the alternatives was the rehabilitation of the existing 200-East coal-fired facility. The other twelve alternatives are combinations of (1) coal- or gas-fueled plants, (2) steam-only or cogeneration facilities, (3) primary or secondary cogeneration of electricity, and (4) public or private ownership.

  14. ARM - Publications: Science Team Meeting Documents: Analysis of

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

    ground-based radiation and cloud measurements from CRYSTAL-FACE Analysis of ground-based radiation and cloud measurements from CRYSTAL-FACE Mather, James Pacific Northwest National Laboratory Comstock, Jennifer Pacific Northwest National Laboratory The Pacific Northwest National Laboratory (PNNL) operates a mobile observing system, the PNNL Atmospheric Remote Sensing Laboratory (PARSL) that is designed to study clouds and aerosols and their effect on atmospheric radiation. The design of

  15. 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.

  16. Underground Test Area Subproject Phase I Data Analysis Task. Volume VII - Tritium Transport Model Documentation Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume VII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the tritium transport model documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  17. Wind energy conversion system analysis model (WECSAM) computer program documentation

    SciTech Connect (OSTI)

    Downey, W T; Hendrick, P L

    1982-07-01

    Described is a computer-based wind energy conversion system analysis model (WECSAM) developed to predict the technical and economic performance of wind energy conversion systems (WECS). The model is written in CDC FORTRAN V. The version described accesses a data base containing wind resource data, application loads, WECS performance characteristics, utility rates, state taxes, and state subsidies for a six state region (Minnesota, Michigan, Wisconsin, Illinois, Ohio, and Indiana). The model is designed for analysis at the county level. The computer model includes a technical performance module and an economic evaluation module. The modules can be run separately or together. The model can be run for any single user-selected county within the region or looped automatically through all counties within the region. In addition, the model has a restart capability that allows the user to modify any data-base value written to a scratch file prior to the technical or economic evaluation. Thus, any user-supplied data for WECS performance, application load, utility rates, or wind resource may be entered into the scratch file to override the default data-base value. After the model and the inputs required from the user and derived from the data base are described, the model output and the various output options that can be exercised by the user are detailed. The general operation is set forth and suggestions are made for efficient modes of operation. Sample listings of various input, output, and data-base files are appended. (LEW)

  18. Documentation and analysis for packaging for surface moisture measurement system 7A containers

    SciTech Connect (OSTI)

    Clem, D.K.

    1996-06-17

    This documentation and analysis for packaging documents that two, procured, carbon steel 5-gal drums meet all applicable U.S.Department of Transportation-7A requirements. One container will be used to transport a 0.009 Ci 252 Cf source and the other to transport a 1.7 Ci Am-Be source to and from various 200 Area tank farms.

  19. Complete Experiment Safety Documentation

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

    Complaint Form Complaint Form The Office of Inspector General (OIG) maintains a Hotline to facilitate the reporting of allegations of fraud, waste, abuse, or mismanagement in U.S. Department of Energy (DOE) programs or operations. To submit an allegation to the OIG, complete the form below. Acknowledgement * I acknowledge that I have read the Office of Inspector General Hotline section of the OIG website regarding issues which should be reported to the OIG, complaint processing, anonymity and

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

    Office of Scientific and Technical Information (OSTI)

    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 TRACEPARCS code system. ...

  1. Battery Lifetime Analysis and Simulation Tool (BLAST) Documentation

    SciTech Connect (OSTI)

    Neubauer, J.

    2014-12-01

    The deployment and use of lithium-ion batteries in automotive and stationary energy storage applications must be optimized to justify their high up-front costs. Given that batteries degrade with use and storage, such optimizations must evaluate many years of operation. As the degradation mechanisms are sensitive to temperature, state-of-charge histories, current levels, and cycle depth and frequency, it is important to model both the battery and the application to a high level of detail to ensure battery response is accurately predicted. To address these issues, the National Renewable Energy Laboratory has developed the Battery Lifetime Analysis and Simulation Tool (BLAST) suite of tools. This suite of tools pairs NREL's high-fidelity battery degradation model with a battery electrical and thermal performance model, application-specific electrical and thermal performance models of the larger system (e.g., an electric vehicle), application-specific system use data (e.g., vehicle travel patterns and driving data), and historic climate data from cities across the United States. This provides highly realistic, long-term predictions of battery response and thereby enables quantitative comparisons of varied battery use strategies.

  2. Sandia Energy - SCADA Documents

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

    Documents Home Stationary Power Safety, Security & Resilience of Energy Infrastructure Grid Modernization Cyber Security for Electric Infrastructure National Supervisory Control...

  3. 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.

  4. 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.

  5. 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.

  6. Exploratory Nuclear Reactor Safety Analysis and Visualization...

    Office of Scientific and Technical Information (OSTI)

    methodologies allows a systematic approach to uncertainty ... Second, we enable exploratory analysis for users, where ... Country of Publication: United States Language: English Subject: ...

  7. 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.

  8. Underground Test Area Subproject Phase I Data Analysis Task. Volume II - Potentiometric Data Document Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume II of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the potentiometric data. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  9. 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 ...

  10. 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.

  11. 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.

  12. 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.

  13. System for the Analysis of Global Energy Markets - Vol. I, Model Documentation

    Reports and Publications (EIA)

    2003-01-01

    Documents the objectives and the conceptual and methodological approach used in the development of projections for the International Energy Outlook. The first volume of this report describes the System for the Analysis of Global Energy Markets (SAGE) methodology and provides an in-depth explanation of the equations of the model.

  14. Safety

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

    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 JLF

  15. 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.

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

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

    documented safety analysis. BACKGROUND On March 11, 2011, the Fukushima Daiichi nuclear power station in Japan was ... of the operating reactors due to the earthquake, a ...

  17. 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.

  18. 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...

  19. Implementation Guide for Use in Developing Documented Safety Analyses to Meet Subpart B of 10 CFR 830

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

    2011-12-19

    Title 10 Code of Federal Regulations (CFR) Part 830, Subpart B, Safety Basis Requirements, requires the contractor responsible for a Department of Energy (DOE) nuclear facility to analyze the facility, the work to be performed, and the associated hazards and to identify the conditions, safe boundaries, and hazard controls necessary to protect workers, the public, and the environment from adverse consequences. Supersedes DOE G 421.1-2.

  20. 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)

  1. Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

    SciTech Connect (OSTI)

    Mahaffy, J. [Pennsylvania State Univ., University Park, PA (United States); Boyack, B.E.; Steinke, R.G. [Los Alamos National Lab., NM (United States)

    1998-05-01

    The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization.

  2. 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

  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. 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.

  5. Data development technical support document for the aircraft crash risk analysis methodology (ACRAM) standard

    SciTech Connect (OSTI)

    Kimura, C.Y.; Glaser, R.E.; Mensing, R.W.; Lin, T.; Haley, T.A.; Barto, A.B.; Stutzke, M.A.

    1996-08-01

    The Aircraft Crash Risk Analysis Methodology (ACRAM) Panel has been formed by the US Department of Energy Office of Defense Programs (DOE/DP) for the purpose of developing a standard methodology for determining the risk from aircraft crashes onto DOE ground facilities. In order to accomplish this goal, the ACRAM panel has been divided into four teams, the data development team, the model evaluation team, the structural analysis team, and the consequence team. Each team, consisting of at least one member of the ACRAM plus additional DOE and DOE contractor personnel, specializes in the development of the methodology assigned to that team. This report documents the work performed by the data development team and provides the technical basis for the data used by the ACRAM Standard for determining the aircraft crash frequency. This report should be used to provide the generic data needed to calculate the aircraft crash frequency into the facility under consideration as part of the process for determining the aircraft crash risk to ground facilities as given by the DOE Standard Aircraft Crash Risk Assessment Methodology (ACRAM). Some broad guidance is presented on how to obtain the needed site-specific and facility specific data but this data is not provided by this document.

  6. 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).

  7. 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

  8. 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.

  9. 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

  10. Safety

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

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

  11. 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.

  12. Development of SRC-I product analysis. Volume 3. Documentation of procedures

    SciTech Connect (OSTI)

    Schweighardt, F.K.; Kingsley, I.S.; Cooper, F.E.; Kamzelski, A.Z.; Parees, D.M.

    1983-09-01

    This section documents the BASIC computer program written to simulate Wilsonville's GC-simulated distillation (GCSD) results at APCI-CRSD Trexlertown. The GC conditions used at APCI for the Wilsonville GCSD analysis of coal-derived liquid samples were described in the SRC-I Quarterly Technical Report, April-June 1981. The approach used to simulate the Wilsonville GCSD results is also from an SRC-I Quarterly Technical Report and is reproduced in Appendix VII-A. The BASIC computer program is described in the attached Appendix VII-B. Analysis of gases produced during coal liquefaction generates key information needed to determine product yields for material balance and process control. Gas samples from the coal process development unit (CPDU) and tubing bombs are the primary samples analyzed. A Carle gas chromatographic system was used to analyze coal liquefaction gas samples. A BASIC computer program was written to calculate the gas chromatographic peak area results into mole percent results. ICRC has employed several analytical workup procedures to determine the amount of distillate, oils, asphaltenes, preasphaltenes, and residue in SRC-I process streams. The ASE procedure was developed using Conoco's liquid column fractionation (LC/F) method as a model. In developing the ASE procedure, ICRC was able to eliminate distillation, and therefore quantify the oils fraction in one extraction step. ASE results were shown to be reproducible within +- 2 wt %, and to yield acceptable material balances. Finally, the ASE method proved to be the least affected by sample composition.

  13. 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 the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical

  14. 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.

  15. 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.

  16. Underground Test Area Subproject Phase I Data Analysis Task. Volume III - Groundwater Recharge and Discharge Data Documentation Package

    SciTech Connect (OSTI)

    1996-10-01

    Volume III of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the data covering groundwater recharge and discharge. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  17. Underground Test Area Subproject Phase I Data Analysis Task. Volume V - Transport Parameter and Source Term Data Documentation Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume V of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the transport parameter and source term data. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  18. 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.

  19. 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 and requirements of reviews; and (5) Provide the above information to DOE organizations, contractors, other government agencies, and interested members of the general public. This PRG was originally published in September 1987. Revision 1, issued in October 1988, added new review sections on quality assurance and penetrations through the containment boundary, along with a few other items. Revision 2 was published October 1999. Revision 3 of this PRG is a complete update, and supersedes Revision 2 in its entirety.

  20. 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.

  1. 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.

  2. TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

    SciTech Connect (OSTI)

    KRIPPS, L.J.

    2006-07-31

    This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  3. 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.

  4. LANL Nuclear Safety Support Services IDIQ Contract

    Broader source: Energy.gov [DOE]

    The scope of this contract will consist of technical support services needed to support the development and implementation of Documented Safety Analysis for EM operations facilities at LANL, such as Technical Area 54 Area G.

  5. 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.

  6. Strategic Safety Goals | Department of Energy

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

    Safety Goals More Documents & Publications Strategic Safety Goals Occupational Safety Performance Trends Development of the Nuclear Safety Information Dashboard - September 2012

  7. 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.

  8. TEC Working Group Topic Groups Rail Archived Documents | Department...

    Office of Environmental Management (EM)

    Archived Documents TEC Working Group Topic Groups Rail Archived Documents ARCHIVED DOCUMENTS PDF icon Inspections Summary Matrix PDF icon TEC Transportation Safety WIPP-PIG Rail...

  9. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect (OSTI)

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

  10. 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.

  11. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    SciTech Connect (OSTI)

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  12. 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.

  13. Electronic document management system analysis report and system plan for the Environmental Restoration Program

    SciTech Connect (OSTI)

    Frappaolo, C.

    1995-09-01

    Lockheed Martin Energy Systems, Inc. (LMES) has established and maintains Document Management Centers (DMCs) to support Environmental Restoration Program (ER) activities undertaken at three Oak Ridge facilities: Oak Ridge National Laboratory, Oak Ridge K-25 Site, Oak Ridge Y-12 Plant; and two sister sites: Portsmouth Gaseous Diffusion Plant in Portsmouth, Ohio, and Paducah Gaseous Diffusion Plant in Paducah, Kentucky. The role of the DMCs is to receive, store, retrieve, and properly dispose of records. In an effort to make the DMCs run more efficiently and to more proactively manage the records` life cycles from cradle to grave, ER has decided to investigate ways in which Electronic Document Management System (EDMS) technologies can be used to redefine the DMCs and their related processes. Specific goals of this study are tightening control over the ER documents, establishing and enforcing record creation and retention procedures, speeding up access to information, and increasing the accessibility of information. A working pilot of the solution is desired within the next six months. Based on a series of interviews conducted with personnel from each of the DMCs, key management, and individuals representing related projects, it is recommended that ER utilize document management, full-text retrieval, and workflow technologies to improve and automate records management for the ER program. A phased approach to solution implementation is suggested starting with the deployment of an automated storage and retrieval system at Portsmouth. This should be followed with a roll out of the system to the other DMCs, the deployment of a workflow-enabled authoring system at Portsmouth, and a subsequent roll out of this authoring system to the other sites.

  14. 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.

  15. NEW - DOE O 420.1 Chg 1, Facility Safety

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

    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, cancels DOE O 420.1C, dated 12-4-12.

  16. Setting clear expectations for safety basis development

    SciTech Connect (OSTI)

    MORENO, M.R.

    2003-05-03

    DOE-RL has set clear expectations for a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (10 CFR 830, Nuclear Safety Rule) which will ensure long-term benefit to Hanford. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development resulting in a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was issued to standardized methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was issued for the evaluation of radiological consequences for accident scenarios often postulated for Hanford. A standard Site Documented Safety Analysis (DSA) detailing the safety management programs was issued for use as a means of compliance with a majority of 3009 Standard chapters. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. As a result of setting expectations and providing safety analysis tools, the four Hanford Site waste management nuclear facilities were able to integrate into one Master Waste Management Documented Safety Analysis (WM-DSA).

  17. FAQS Reference Guide - Senior Technical Safety Manager | Department of

    Energy Savers [EERE]

    Energy Senior Technical Safety Manager FAQS Reference Guide - Senior Technical Safety Manager This reference guide has been developed to address the competency statements in the March 2013 edition of DOE-Standard (STD)-1175-2013, Senior Technical Safety Manager Functional Area Qualification Standard. PDF icon Senior Technical Safety Manager Qualification Standard Reference Guide, October 2013 More Documents & Publications FAQS Gap Analysis Qualification Card - Senior Technical Safety

  18. Enforcement Documents | Department of Energy

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

    Enforcement » Enforcement Documents Enforcement Documents RSS May 9, 2016 Consent Order, Battelle Energy Alliance, LLC Worker Safety and Health Enforcement Consent Order issued to Battelle Energy Alliance, LLC relating to an arc flash event at the Idaho National Laboratory. April 28, 2016 Savannah River Nuclear Solutions, LLC, Consent Order NCO-2016-01 Nuclear Safety Enforcement Consent Order issued to Savannah River Nuclear Solutions, LLC relating to nuclear criticality safety infractions that

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Other Documents

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

    since 2009 5 U.S.C. § 552 10 CFR 1004 Other Documents Transuranic Waste Defense Determination Approval Form

  4. WIPP Documents

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

    Large file size alert This symbol means the document may be a large file size. Environmental Monitoring Documents about environmental monitoring at WIPP, including annual site environmental reports Environmental Protection Agency Certification Documents about WIPP's EPA certification Federal Regulations Standards and compliance criteria for WIPP Hazardous Waste Facility Permit Documents related to the state of New Mexico's oversight of WIPP under the authority of the Resource Conservation and

  5. 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.

  6. FAQS Gap Analysis Qualification Card – Nuclear Safety Specialist

    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.

  7. 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.

  8. 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.

  9. 2011 Annual Workforce Analysis and Staffing Plan Report - Oak...

    Energy Savers [EERE]

    safety assurance. PDF icon 2011 Annual Workforce Analysis and Staffing Plan Report - ORO More Documents & Publications 2010 Annual Workforce Analysis and Staffing Plan Report -...

  10. Guidance Related to Analysis of Impacts to Workers in National Policy Act (NEPA) Documentation

    Broader source: Energy.gov [DOE]

    The purpose of this memorandum is to provide general guidance regarding the analysis of impacts to workers in DOE environmental impact statements (EISs) and environmental assessments (EAs) in order...

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. System for the Analysis of Global Energy Markets - Vol. II, Model Documentation

    Reports and Publications (EIA)

    2003-01-01

    The second volume provides a data implementation guide that lists all naming conventions and model constraints. In addition, Volume 1 has two appendixes that provide a schematic of the System for the Analysis of Global Energy Markets (SAGE) structure and a listing of the source code, respectively.

  17. Hazard baseline documentation

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    This DOE limited technical standard establishes uniform Office of Environmental Management (EM) guidance on hazards baseline documents that identify and control radiological and nonradiological hazards for all EM facilities. It provides a road map to the safety and health hazard identification and control requirements contained in the Department`s orders and provides EM guidance on the applicability and integration of these requirements. This includes a definition of four classes of facilities (nuclear, non-nuclear, radiological, and other industrial); the thresholds for facility hazard classification; and applicable safety and health hazard identification, controls, and documentation. The standard applies to the classification, development, review, and approval of hazard identification and control documentation for EM facilities.

  18. 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.

  19. Cost-Effectiveness Analysis of the 2009 and 2012 IECC Residential Provisions – Technical Support Document

    SciTech Connect (OSTI)

    Mendon, Vrushali V.; Lucas, Robert G.; Goel, Supriya

    2012-12-04

    This analysis was conducted by Pacific Northwest National Laboratory (PNNL) in support of the U.S. Department of Energy’s (DOE) Building Energy Codes Program (BECP). DOE supports the development and adoption of efficient residential and commercial building energy codes. These codes set the minimum requirements for energy efficient building design and construction and ensure energy savings on a national level. This analysis focuses on one and two family dwellings, townhomes, and low-rise multifamily residential buildings. For these buildings, the basis of the energy codes is the International Energy Conservation Code (IECC). This report does not address commercial and high-rise residential buildings, which reference ANSI/ASHRAE/IES Standard 90.1.

  20. 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.

  1. Hanford Generic Interim Safety Basis

    SciTech Connect (OSTI)

    Lavender, J.C.

    1994-09-09

    The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  2. Tracking and Analysis Framework (TAF) model documentation and user`s guide

    SciTech Connect (OSTI)

    Bloyd, C.; Camp, J.; Conzelmann, G.

    1996-12-01

    With passage of the 1990 Clean Air Act Amendments, the United States embarked on a policy for controlling acid deposition that has been estimated to cost at least $2 billion. Title IV of the Act created a major innovation in environmental regulation by introducing market-based incentives - specifically, by allowing electric utility companies to trade allowances to emit sulfur dioxide (SO{sub 2}). The National Acid Precipitation Assessment Program (NAPAP) has been tasked by Congress to assess what Senator Moynihan has termed this {open_quotes}grand experiment.{close_quotes} Such a comprehensive assessment of the economic and environmental effects of this legislation has been a major challenge. To help NAPAP face this challenge, the U.S. Department of Energy (DOE) has sponsored development of an integrated assessment model, known as the Tracking and Analysis Framework (TAF). This section summarizes TAF`s objectives and its overall design.

  3. 2007 Wholesale Power Rate Case Initial Proposal : Risk Analysis Study Documentation.

    SciTech Connect (OSTI)

    United States. Bonneville Power Administration.

    2005-11-01

    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-E-BPA-03); the RAM2007 is the computer model being used to calculate rates (see Wholesale Power Rate Development Study, WP-07-E-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-E-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-E-BPA-01, various revenues from the Revenue Forecast component of the Wholesale Power Rate Development Study, WP-07-E-BPA-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 ToolKit Model to develop the risk mitigation package that achieves BPA's 92.6 percent TPP standard. The processes and interaction between each of the models and studies are depicted in Graph 1.

  4. 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 ToolKit Model to develop the risk mitigation package that achieves BPA's 92.6 percent TPP standard. The processes and interaction between each of the models and studies are depicted in Graph 1.

  5. Implementation Guide for Use in Developing Technical Safety Requirements

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

    2010-11-03

    This Guide provides elaboration for the content of TSRs. Section 10 CFR 830.205 of the Nuclear Safety Management rule, requires Department of Energy (DOE) contractors responsible for category 1, 2, and 3 DOE nuclear facilities to develop Technical Safety Requirements (TSRs). These TSRs identify the limitations to each DOE owned, contractor operated nuclear facility based on the documented safety analysis (DSA) and any additional safety requirements established for the facility. Cancels DOE G 423.1-1.

  6. Implementation Guide for Use in Developing Technical Safety Requirements

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

    2010-11-03

    This Guide provides elaboration for the content of TSRs. Section 10 CFR 830.205 of the Nuclear Safety Management rule, requires Department of Energy (DOE) contractors responsible for category 1, 2, and 3 DOE nuclear facilities to develop Technical Safety Requirements (TSRs). These TSRs identify the limitations to each DOE owned, contractor operated nuclear facility based on the documented safety analysis (DSA) and any additional safety requirements established for the facility. Does not cancel other directives.

  7. Untitled Document

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

    For questions on the Environmental Assessment for Conducting Astrophysics and Other Basic Science Experiments at the WIPP Site, email Harold.Johnson@ or call (505) 234-7349. Environmental Assessment for Conducting Astrophysics and Other Basic Science Experiments at the WIPP Site Final - January, 2001 This document has been provided to you in PDF format. Please install Adobe Acrobat Reader before accessing these documents. Cover Sheet, Table of Contents, List of Tables, List of Figures, and

  8. Document Delivery

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

    Services » Document Delivery Document Delivery The Research Library can order journal articles, conference papers and book chapters not found in the Library collection for LANL staff. The service is provided by our vendor, Reprints Desk, who will mail the article to you as a PDF. Easily retrieved articles may be delivered the same day, while harder-to-find material can take several days/weeks. This service is limited to LANL employees with active Z numbers. Requesting Materials Use the Request

  9. Supplement analysis for continued operation of Lawrence Livermore National Laboratory and Sandia National Laboratories, Livermore. Volume 2: Comment response document

    SciTech Connect (OSTI)

    1999-03-01

    The US Department of Energy (DOE), prepared a draft Supplement Analysis (SA) for Continued Operation of Lawrence Livermore National Laboratory (LLNL) and Sandia National Laboratories, Livermore (SNL-L), in accordance with DOE`s requirements for implementation of the National Environmental Policy Act of 1969 (NEPA) (10 Code of Federal Regulations [CFR] Part 1021.314). It considers whether the Final Environmental Impact Statement and Environmental Impact Report for Continued Operation of Lawrence Livermore National Laboratory and Sandia National Laboratories, Livermore (1992 EIS/EIR) should be supplement3ed, whether a new environmental impact statement (EIS) should be prepared, or no further NEPA documentation is required. The SA examines the current project and program plans and proposals for LLNL and SNL-L, operations to identify new or modified projects or operations or new information for the period from 1998 to 2002 that was not considered in the 1992 EIS/EIR. When such changes, modifications, and information are identified, they are examined to determine whether they could be considered substantial or significant in reference to the 1992 proposed action and the 1993 Record of Decision (ROD). DOE released the draft SA to the public to obtain stakeholder comments and to consider those comments in the preparation of the final SA. DOE distributed copies of the draft SA to those who were known to have an interest in LLNL or SNL-L activities in addition to those who requested a copy. In response to comments received, DOE prepared this Comment Response Document.

  10. Controlling OSHA PSM documentation

    SciTech Connect (OSTI)

    Person, J.S. [Warren-Forthought, Inc., Angleton, TX (United States)

    1997-06-01

    At present, most companies in the hazardous materials process industries are knowledgeable in the requirements of the OSHA Process Safety Management (PSM) standard (29 CFR 1910.119), issued in May of 1992. The current question facing many of these companies is no longer ``How does one comply?`` but rather, ``What does one do with all the documentation generated from the OSHA PSM Standard?``. Using an Electronic Performance Support System (EPSS) for delivering and controlling OSHA PSM documentation will be discussed. This paper will address the important components of an EPSS and successful case studies.

  11. 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 intermediate system is possible, raising questions about the effectiveness of the intermediate decay heat removal systems in the design that was evaluated. Existing tools do not predict flow stagnation. This work has demonstrated that with a proper coupling approach, a high-fidelity CFD tool can be used to resolve the important flow and temperature distributions throughout a plant while still maintaining the whole-plant safety analysis capabilities of a systems analysis code.

  12. DOCUMENT RELEASE FORM

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

    6992 (2) Revision Number: 0 1(3) Effective Date: 9.- -, (4) Document Type: [] Digital Image DHard copy (a) Number of pages (including the DIRF) or Z PDF ~ zVideo number of digital images q- (5) Release Type [DNew E71 Cancel 11 Page Change El Complete Revision (6) Document Title: Analysis of Power Delivery to the WTP to Achieve a Minimum of 70MW (76.4MVA) (7) Change/Release Initial Release of Document Description: (8) Change Initial Release of Document Justification: (9) Associated (a) Structure

  13. Ames Laboratory Forms and Documents | The Ames Laboratory

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

    Ames Laboratory Forms and Documents Document Type Bulletin Charter Form Guide Handout Lessons Learned Manual Material Safety Data Sheet Memo Misc Packet Plan Policy Procedure...

  14. Safety Design Strategy RM | Department of Energy

    Energy Savers [EERE]

    Safety Design Strategy RM Safety Design Strategy RM The SDS Review Module (RM) is a tool that assists DOE federal project review teams in evaluating the adequacy of the conceptual safety design strategy documentation package (Conceptual Safety Design Report) prior to CD-1 approval and the adequacy of the preliminary safety design strategy documentation packaged (Preliminary Safety Design Report) prior to CD- 2 approval. PDF icon Safety Design Strategy RM More Documents & Publications Safety

  15. RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    KOZLOWSKI, S.D.

    2007-05-30

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

  16. 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.

  17. Appendix G Supporting Documentation

    Office of Legacy Management (LM)

    Supporting Documentation for Focused Feasibility Study This page intentionally left blank Contents Section Ground Water Models for Alternatives 3 and 4 ........................................................................ G .O Human Health Risk Assessment Calculations ........................................................................... G2.0 Cost Analysis ........................................................................................................................... G3.0 This

  18. Untitled Document

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

    Remote-Handled Transuranic Waste Study - Main Menu DOE/CAO 95-1095 REMOTE-HANDLED TRANSURANIC WASTE STUDY U.S. Department of Energy Carlsbad Area Office Carlsbad, New Mexico October 1995 This document has been reproduced directly from the best possible copy. It is available to DOE and DOE contractors at the following address: Office of Scientific and Technical Information P. O. Box 62 Oak Ridge, TN 37831 For prices contact (615) 576-8401 The local WIPP contact for this document is George

  19. 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 and 1988. All of the reports reviewed were the regulator’s assessment of the PSR findings rather than the original PSR report, and all but one were English translations from the original language. In these reviews, it was found that most of the countries base their regulatory guidance to some extent (and often to a large extent) on U.S. design codes and standards, NRC regulatory guidance, and U.S. industry guidance. In addition, many of the observed operational technical issues and OpE events reported for U.S. reactors are also cited in the PSR reports. The PSR reports also identified a number of potential technical material/component performance issues and OpE events that are not commonly reported for U.S. plants.

  20. 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.

  1. Presidential Documents

    Energy Savers [EERE]

    425 Federal Register / Vol. 80, No. 23 / Wednesday, February 4, 2014 / Presidential Documents Executive Order 13690 of January 30, 2015 Establishing a Federal Flood Risk Management Standard and a Process for Further Soliciting and Considering Stakeholder Input By the authority vested in me as President by the Constitution and the laws of the United States of America, and in order to improve the Nation's resilience to current and future flood risk, I hereby direct the following: Section 1.

  2. Coiled Tubing Safety Manual

    SciTech Connect (OSTI)

    Crow, W.

    1999-04-06

    This document addresses safety concerns regarding the use of coiled tubing as it pertains to the preservation of personnel, environment and the wellbore.

  3. Untitled Document

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    | Gasoline | Diesel | Propane | Natural Gas | Electricity | Coal | Nuclear Renewables | Alternative Fuels | Prices | States | International | Country Analysis Briefs Environment...

  4. Enron Documents

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

    f Enron Documents 1994 U.S. DEPARTMENT OF ENERGY 13S. 12 CORRESPONDENCE CONTROL FORM 1) OFFICE OF THE EXECUTIVE SECRETARIAT 8CTIVITY ADD DO NOT DETACH FROM ORIGINAL CORRESPONDENCE 0/03/Y4 10: O0 5OURCECODE PM PUBLIC MAIL :?EC INT: CONTROL NO: ES94-012853 ,ATE CORR: 04/25/94 DATE RECD: 05/02/94 DATE CNTRL: 05/03/94 DATE DUE: NONE :TTER: X MEMO: TWX: _ OTHER: TO: SECY: X DEP SEC: _ UN SEC: OTHER: )M:LAY,KENNETH L \ TX O REMARKS: RELATES TO ES94-009882. i ENRON CORP. J: PUBLIC INFORMATION NATIONAL

  5. AFFECTED DOCUMENT

    Office of Legacy Management (LM)

    ADMIN RCD _ COMM REF _ RESPONSE TRACKING INFORMAnON I lOWED BY: (ORG) I TARGET DATE I I I CLOSING CCN COMPL. DATE I I CLOSING REF I I lOWED BY: (ORG) I I I I I TARGET DATE I CLOSING CCN COMPL DATE CLOSING REF TECHNICAL SERVICES DIVISION (TSD) BACKFIT (Documents dated prior to 1 November 1988) FUSRAP COMMUNICATIONS DISTRIBUTION DOEIORO TECHNICAL SERVICES DIVISION (CE-53) BECHTEL NATIONAl.. INC. - JOB 14501 AI$Wl E.e... COMM DATE b~5i=-'/T-'-I'''' ADDR CODE I I I I CLOSES CCN WBS 1/~ .. :2 O~

  6. AFFECTED DOCUMENT

    Office of Legacy Management (LM)

    rr-osams ADMIN RCD _ RESPONSE TRACKING INFORMATION I 'OWED BY: (ORG) I TARGET DATE { I CLOSING CCN COMPL. DATE { CLOSING REF I I lOWED BY: (ORG) { { I { { TARGET DATE I CLOSING CCN COMPL DATE CLOSING REF TECHNICAL SERVICES DIVISION (TSD) BACKFIT (Documents dated prior to 1 November 1988) FUSRAP COMMUNICATIONS DISTRIBUTION DOEIORO TECHNICAL SERVICES DIVISION (CE*53) BECHTEL NATIONAL, INC. -JOB 14501 _-....Iome....:.-R..%.JI&~S::....:::S.::...- TO_---.,!5~Z~IJ~u..~ COMM DATE &6 tlJ t.e

  7. 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).

  8. 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.

  9. Assessment Documents | Department of Energy

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

    Safety Analysis April 10, 2015 Enterprise Assessments Review, Pantex Plant 2014 Full Participation Exercise - April 2015 Review of the Pantex Plant 2014 Full Participation...

  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

    During the Engineering Design Activity of the International Thermonuclear Experimental Reactor (ITER), the MELCOR 1.8.2 code was selected as one of several codes to be used to perform ITER safety analyses [1]. MELCOR was chosen because it has the capability of predicting coolant pressure, temperature, mass flow rate, and radionuclide and aerosol transport in nuclear facilities and reactor cooling systems. MELCOR can also predict structural temperatures (e.g. first wall, blanket, divertor, and vacuum vessel) resulting from energy produced by radioactive decay heat and/or chemical reactions (oxidation). The Idaho National Laboratory (INL) Fusion Safety Program (FSP) made fusion specific modifications to the MELCOR 1.8.2 code [2-6], including models for water freezing, air condensation, beryllium, carbon, and tungsten oxidation in steam and air environments, flow boiling in coolant loops, and radiation in enclosures, that allowed MELCOR to assess the thermal hydraulic response of ITER cooling systems and the transport of radionuclides as aerosols during accident conditions. Recently, the ITER International Organization (IO) used a “pedigreed” version of MELCOR 1.8.2 [7] to perform accident analyses for ITER’s “Rapport Préliminaire de Sûreté” (Report Preliminary on Safety - RPrS). The MELCOR thermal-hydraulics code [8] is currently under development at the Sandia National Laboratory (SNL) for the US Nuclear Regulatory Commission (NRC). MELCOR is used to model the progression of severe accidents in light water fission reactors. Because MELCOR has undergone many improvements between version 1.8.2 and 1.8.6, the INL FSP decided to introduce fusion modifications into MELCOR 1.8.6, and thereby produce a version of MELCOR 1.8.6 with similar capabilities to the pedigreed version of MELCOR 1.8.2 used for the ITER RPrS. We have applied this version of MELCOR 1.8.6 to the same set of problems used in the MELCOR 1.8.2 pedigree analysis [7]. Section 2 describes a non-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. Untitled Document

    Gasoline and Diesel Fuel Update (EIA)

    Revision to the Analysis of the Impacts of the Clean Power Plan May 27, 2015 Minor changes to Table 7. Nuclear power projections in three cases...

  12. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  13. FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

    SciTech Connect (OSTI)

    GARVIN, L J; JENSEN, M A

    2004-04-13

    This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

  14. 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.

  15. 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.

  16. 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.

  17. ARM - Conference Documents

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

    govPublicationsConference Documents Publications Journal Articles Conference Documents Program Documents Technical Reports Publications Database Public Information Materials Image...

  18. 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.

  19. Scanned Document

    Office of Legacy Management (LM)

    EPA-402-R-04-006 Environmental Protection Indoor Air November 2004 Agency Washington, DC 20460 Annual Water Sampling and Analysis, Calendar Year 2004: SHOAL Test Site Area FAULTLESS Test Site Area RULISON Test Site Area RIO BLANCO Test Site Area GASBUGGY Test Site Area GNOME Test Site Area Annual Water Sampling and Analysis, Calendar Year 2004 SHOAL Test Site Area FAULTLESS Test Site Area RULISON Test Site Area RIO BLANCO Test Site Area GASBUGGY Test Site Area GNOME Test Site Area Max G. Davis

  20. Safety Planning Guidance for Hydrogen and Fuel Cell Projects

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

    ... of Safety Vulnerabilities (ISV) o Risk Reduction Plan o Operating Procedures - ... o Project Safety Documentation 4. Communication Plan o Employee Training o Safety ...

  1. 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 the Office of Safety Management and Operations. Differences between the Model 9975 Package and the Model 9977 Package include: (1) The lead shield present in the Model 9975 Package is absent in the Model 9977 Package; (2) The Model 9975 Package has eight allowable contents, while the Model 9977 Package has a single allowable content. (3) The 6CV of the Model 9977 Package is similar in design to the outer Containment Vessel of the Model 9975 Package that also incorporates a 5-inch Containment Vessel as the inner Containment Vessel. (4) The Model 9975 Package uses a Celotex{reg_sign}-based impact limiter while the Model 9977 Package uses Last-A-Foam{reg_sign}, a polyurethane foam, for the impact limiter. (5) The Model 9975 Package has two Containment Vessels, while the Model 9977 Package has a single Containment Vessel.

  2. Nuclear Safety Enforcement Documents | Department of Energy

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

    of an Irradiated Pellet at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory August 18, 1999 Preliminary Notice of Violation,...

  3. Nuclear Safety Enforcement Documents | Department of Energy

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

    LLC, related to Quality Assurance Violations and Deficiencies at the Idaho National Engineering and Environmental Laboratory October 3, 2000 Enforcement Letter, Westinghouse...

  4. Preparation of Nonreactor Nuclear Facility Documented Safety...

    Energy Savers [EERE]

    ......... 21 3.2.4.3 Chemical Source Term and Consequence ... and hazardous materials by mechanical, chemical, or other techniques to achieve a stated ...

  5. Nuclear Safety Enforcement Documents | Department of Energy

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

    Idaho Technologies Company related to Unplanned Internal Radiation Exposures at the Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory,...

  6. Nuclear Safety Enforcement Documents | Department of Energy

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

    Company, LLC related to Fire Protection System Surveillances at the Rocky Flats Environmental Technology Site, July 21, 1998 July 21, 1998 Enforcement Letter,...

  7. Nuclear Safety Enforcement Documents | Department of Energy

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

    Flats Environmental Technology Site August 9, 2004 Preliminary Notice of Violation, Fluor Fernald, Inc. - EA-2004-07 Issued to Fluor Fernald, Inc., related to Respiratory...

  8. Nuclear Safety Enforcement Documents | Department of Energy

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

    Hanford Waste Treatment and Immobilization Plant Project April 13, 2010 Consent Order, Parsons Infrastructure & Technology Group, Inc. - NCO-2010-02 Issued to Parsons...

  9. Nuclear Safety Enforcement Documents | Department of Energy

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

    Preliminary Notice of Violation, BWXT Y-12 - EA-2003-03 Issued to BWXT Y-12, related to Welding Inspection Deficiencies at the Y-12 National Security Complex, June 4, 2003...

  10. 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

  11. 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).

  12. 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.

  13. 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.

  14. 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.

  15. Untitled Document

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

    Revision to the Analysis of the Impacts of the Clean Power Plan May 27, 2015 Minor changes to Table 7. Nuclear power projections in three cases

    3 Meeting of the American Statistical Association Committee on Energy Statistics and the Energy Information Administration This is an unedited Transcript of the Fall Meeting of the Energy Information Administration and the American Statistical Association Committee on Energy Statistics. The meeting took place October 16 and 17, 2003 in the Forrestal

  16. Safety and Security Enforcement Coordinator Handbook | Department...

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

    Coordinator Handbook Safety and Security Enforcement Coordinator Handbook April 2015 This handbook is a companion document to the Enforcement Process Overview (EPO). The Safety and ...

  17. Security, Emergency Planning & Safety Records | Department of...

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

    Security, Emergency Planning & Safety Records Security, Emergency Planning & Safety Records PDF icon ADM 180.pdf More Documents & Publications ADMINISTRATIVE RECORDS SCHEDULE 18: ...

  18. Hydrogen Safety Knowledge Tools | Department of Energy

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

    scs_04_fassbender.pdf More Documents & Publications What Can We Learn from Hydrogen Safety Event Databases? H2 Safety Snapshot, Vol. 1, Issue 1, April 2009

  19. Safety and Security Enforcement Coordinator Handbook | Department...

    Energy Savers [EERE]

    Safety and Security Enforcement Coordinator Handbook Safety and Security Enforcement Coordinator Handbook April 2015 This handbook is a companion document to the Enforcement...

  20. Documentation - Laboratory for Laser Energetics

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

    VII System Description Chapter 1: System Overview Chapter 2: Laser Sources (final draft material) Chapter 3: Laser Amplifiers Chapter 4: Power Conditioning Chapter 5: Optomechanical System Chapter 6: Laser Diagnostics Chapter 7: Experimental System Chapter 8: Control System Chapter 9: Facility and Safety Interlocks Comments Address document comments, questions and corrections to the OMEGA EP Associate Laser Facility Manager

  1. Melter Disposal Strategic Planning Document

    SciTech Connect (OSTI)

    BURBANK, D.A.

    2000-09-25

    This document describes the proposed strategy for disposal of spent and failed melters from the tank waste treatment plant to be built by the Office of River Protection at the Hanford site in Washington. It describes program management activities, disposal and transportation systems, leachate management, permitting, and safety authorization basis approvals needed to execute the strategy.

  2. Cray XE Documentation

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

    Cray XE Documentation Cray XE Documentation Last edited: 2014-01-22 15:12:50

  3. CRITICALITY SAFETY CONTROLS AND THE SAFETY BASIS AT PFP

    SciTech Connect (OSTI)

    Kessler, S

    2009-04-21

    With the implementation of DOE Order 420.1B, Facility Safety, and DOE-STD-3007-2007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities', a new requirement was imposed that all criticality safety controls be evaluated for inclusion in the facility Documented Safety Analysis (DSA) and that the evaluation process be documented in the site Criticality Safety Program Description Document (CSPDD). At the Hanford site in Washington State the CSPDD, HNF-31695, 'General Description of the FH Criticality Safety Program', requires each facility develop a linking document called a Criticality Control Review (CCR) to document performance of these evaluations. Chapter 5, Appendix 5B of HNF-7098, Criticality Safety Program, provided an example of a format for a CCR that could be used in lieu of each facility developing its own CCR. Since the Plutonium Finishing Plant (PFP) is presently undergoing Deactivation and Decommissioning (D&D), new procedures are being developed for cleanout of equipment and systems that have not been operated in years. Existing Criticality Safety Evaluations (CSE) are revised, or new ones written, to develop the controls required to support D&D activities. Other Hanford facilities, including PFP, had difficulty using the basic CCR out of HNF-7098 when first implemented. Interpretation of the new guidelines indicated that many of the controls needed to be elevated to TSR level controls. Criterion 2 of the standard, requiring that the consequence of a criticality be examined for establishing the classification of a control, was not addressed. Upon in-depth review by PFP Criticality Safety staff, it was not clear that the programmatic interpretation of criterion 8C could be applied at PFP. Therefore, the PFP Criticality Safety staff decided to write their own CCR. The PFP CCR provides additional guidance for the evaluation team to use by clarifying the evaluation criteria in DOE-STD-3007-2007. In reviewing documents used in classifying controls for Nuclear Safety, it was noted that DOE-HDBK-1188, 'Glossary of Environment, Health, and Safety Terms', defines an Administrative Control (AC) in terms that are different than typically used in Criticality Safety. As part of this CCR, a new term, Criticality Administrative Control (CAC) was defined to clarify the difference between an AC used for criticality safety and an AC used for nuclear safety. In Nuclear Safety terms, an AC is a provision relating to organization and management, procedures, recordkeeping, assessment, and reporting necessary to ensure safe operation of a facility. A CAC was defined as an administrative control derived in a criticality safety analysis that is implemented to ensure double contingency. According to criterion 2 of Section IV, 'Linkage to the Documented Safety Analysis', of DOESTD-3007-2007, the consequence of a criticality should be examined for the purposes of classifying the significance of a control or component. HNF-PRO-700, 'Safety Basis Development', provides control selection criteria based on consequence and risk that may be used in the development of a Criticality Safety Evaluation (CSE) to establish the classification of a component as a design feature, as safety class or safety significant, i.e., an Engineered Safety Feature (ESF), or as equipment important to safety; or merely provides defense-in-depth. Similar logic is applied to the CACs. Criterion 8C of DOE-STD-3007-2007, as written, added to the confusion of using the basic CCR from HNF-7098. The PFP CCR attempts to clarify this criterion by revising it to say 'Programmatic commitments or general references to control philosophy (e.g., mass control or spacing control or concentration control as an overall control strategy for the process without specific quantification of individual limits) is included in the PFP DSA'. Table 1 shows the PFP methodology for evaluating CACs. This evaluation process has been in use since February of 2008 and has proven to be simple and effective. Each control identified i

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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 Packagings use Celotex{trademark} for insulation and as impact limiters. To provide a historical perspective, it is noted that the Model 9975-96 Packaging 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 Model 9978 Package has six Content Envelopes: C.1 ({sup 238}Pu Heat Sources), C.2 ( Pu/U Metals), C.3 (Pu/U Oxides, Reserved), C.4 (U Metal or Alloy), C.5 (U Compounds), and C.6 (Samples and Sources). Per 10 CFR 71.59 (Code of Federal Regulations), the value of N is 50 for the Model 9978 Package leading to a Criticality Safety Index (CSI) of 1.0. The Transport Index (TI), based on dose rate, is calculated to be a maximum of 4.1.

  9. 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.

  10. ARM - XDC Documentation

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

    Documentation External Data Center Order Data Description of External Data Streams Data Viewers and Plots (selected data sets) XDC Documentation XDC Documentation BNL's ARM Related...

  11. ARM - Data Documentation

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

    Documentation Policies, Plans, Descriptions Data Documentation Home Data Sharing and Distribution Policy Data Management and Documentation Plan Data Product Registration and ...

  12. 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.

  13. Cold Vacuum Drying Facility hazard analysis report

    SciTech Connect (OSTI)

    Krahn, D.E.

    1998-02-23

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) hazard analysis to support the CVDF phase 2 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 Reports, and implements the requirements of US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports.

  14. Review and Approval of Nuclear Facility Safety Basis and Safety...

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

    104-2014, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Documents by Website Administrator This Standard describes a framework and the criteria to be...

  15. Supervisors` orientation to occupational safety in DOE

    SciTech Connect (OSTI)

    1993-10-01

    This document presents OSHA regulations, safety and health guidelines pertinent to DOE and the first-line supervisor.

  16. Occupational Safety Performance | Department of Energy

    Energy Savers [EERE]

    Safety Performance More Documents & Publications Acquisition and Project Management Continuous Improvement Presentation DOE ZERH Webinar: Efficient Hot Water Distribution I:...

  17. 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

    The Department of Energy has assigned to Sandia National Laboratories the responsibility of producing a Safety Analysis Report (SAR) for the plutonium-dioxide fueled Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) proposed to be used in the Mars Science Laboratory (MSL) mission. The National Aeronautic and Space Administration (NASA) is anticipating a launch in fall of 2009, and the SAR will play a critical role in the launch approval process. As in past safety evaluations of MMRTG missions, a wide range of potential accident conditions differing widely in probability and seventy must be considered, and the resulting risk to the public will be presented in the form of probability distribution functions of health effects in terms of latent cancer fatalities. The basic descriptions of accident cases will be provided by NASA in the MSL SAR Databook for the mission, and on the basis of these descriptions, Sandia will apply a variety of sophisticated computational simulation tools to evaluate the potential release of plutonium dioxide, its transport to human populations, and the consequent health effects. The first step in carrying out this project is to evaluate the existing computational analysis tools (computer codes) for suitability to the analysis and, when appropriate, to identify areas where modifications or improvements are warranted. The overall calculation of health risks can be divided into three levels of analysis. Level A involves detailed simulations of the interactions of the MMRTG or its components with the broad range of insults (e.g., shrapnel, blast waves, fires) posed by the various accident environments. There are a number of candidate codes for this level; they are typically high resolution computational simulation tools that capture details of each type of interaction and that can predict damage and plutonium dioxide release for a range of choices of controlling parameters. Level B utilizes these detailed results to study many 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.

  18. 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 and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations.

  19. INTERPRETATION REGARDING EXEMPTION RELIEF UNDER 10 C.F.R. PART 820, SUBPART E, EXEMPTION RELIEF, AND NON-COMPLIANT ‘DOCUMENTED SAFETY ANALYSES’ SUBJECT TO 10 C.F.R. PART 830, NUCLEAR SAFETY MANAGEMENT, SUBPART B, SAFETY BASIS REQUIREMENTS

    Broader source: Energy.gov [DOE]

    The following document is the Office of General Counsel (GC) interpretation regarding exemption relief pursuant to 10 C.F.R. Part 820, Procedural Rules for DOE Nuclear Activities, Subpart E,...

  20. CRAD, NNSA- Criticality Safety (CS)

    Office of Energy Efficiency and Renewable Energy (EERE)

    CRAD for Criticality Safety (CS). Criteria Review and Approach Documents (CRADs) that can be used to conduct a well-organized and thorough assessment of elements of safety and health programs.

  1. CRAD, NNSA- Safety Basis (SB)

    Broader source: Energy.gov [DOE]

    CRAD for Safety Basis (SB). Criteria Review and Approach Documents (CRADs) that can be used to conduct a well-organized and thorough assessment of elements of safety and health programs.

  2. 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.

  3. 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.

  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 thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical oehavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.« less

  5. Document | Department of Energy

    Energy Savers [EERE]

    Document Document Extracted Pages PDF icon Document More Documents & Publications EIS-0396: Advance Notice of Intent o Prepare an Environmental Impact Statement EIS-0396: Notice of Intent to Prepare a Programmatic Environmental Impact Statement Meeting Minutes: February 20-21, 2007

    Document Document Extracted Pages PDF icon Document More Documents & Publications ADR Policy Federal Register Microsoft Word - ADR Revised Policy82508Reformatted.doc 2009 ECR FINAL REPORT 2010

  6. High-level waste tank farm set point document

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-15

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

  7. Facility Disposition Safety Strategy RM

    Broader source: Energy.gov [DOE]

    The Facility Disposition Safety Strategy (FDSS) Review Module is a tool that assists DOE federal project review teams in evaluating the adequacy of the facility documentation, preparations or...

  8. Fluor Daniel Hanford contract standards/requirements identification document

    SciTech Connect (OSTI)

    Bennett, G.L.

    1997-04-24

    This document, the Standards/Requirements Identification Document (S/RID) for the Fluor Daniel Hanford Contract, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment.

  9. Guiding Documents and Links | Department of Energy

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

    Guiding Documents and Links Guiding Documents and Links Federal Executive Order 13653, Preparing the United States for the Impacts of Climate Change Executive Order 13693, Planning for Federal Sustainability in the Next Decade DOE DOE Orders DOE Order 430.1B - Real Property and Asset Management DOE Order 436.1 - Departmental Sustainability DOE Policy 450.4A - Integrated Safety Management Policy 2014 DOE Strategic Sustainability Performance Plan LM Documents EMS Description LM Environmental

  10. 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.

  11. 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.

  12. Job Safety

    Office of Environmental Management (EM)

    Job Analysis Template Job Analysis Template Job analysis is the foundation for all assessment and section decisions. This standardized document allows Shared Service Centers (SSCs)/Servicing Human Resources Offices (SHROs) and selecting officials to identify the best person for the job. A sufficient job analysis provides a way to develop this understanding by examining the tasks performed in a job, the competencies required to perform those tasks, and the connection between the tasks and

  13. 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.

  14. 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.

  15. Safety Basis Report

    SciTech Connect (OSTI)

    R.J. Garrett

    2002-01-14

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

  16. Decommissioning Documents | Department of Energy

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

    Decommissioning Documents PDF icon Decommissioning Documents More Documents & Publications Decommissioning Benchmarking Study Final Report Decommissioning Handbook ...

  17. 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.

  18. 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.

  19. Technical support document: Energy conservation standards for consumer products: Dishwashers, clothes washers, and clothes dryers including: Environmental impacts; regulatory impact analysis

    SciTech Connect (OSTI)

    Not Available

    1990-12-01

    The Energy Policy and Conservation Act as amended (P.L. 94-163), establishes energy conservation standards for 12 of the 13 types of consumer products specifically covered by the Act. The legislation requires the Department of Energy (DOE) to consider new or amended standards for these and other types of products at specified times. This Technical Support Document presents the methodology, data and results from the analysis of the energy and economic impacts of standards on dishwashers, clothes washers, and clothes dryers. The economic impact analysis is performed in five major areas: An Engineering Analysis, which establishes technical feasibility and product attributes including costs of design options to improve appliance efficiency. A Consumer Analysis at two levels: national aggregate impacts, and impacts on individuals. The national aggregate impacts include forecasts of appliance sales, efficiencies, energy use, and consumer expenditures. The individual impacts are analyzed by Life-Cycle Cost (LCC), Payback Periods, and Cost of Conserved Energy (CCE), which evaluate the savings in operating expenses relative to increases in purchase price; A Manufacturer Analysis, which provides an estimate of manufacturers' response to the proposed standards. Their response is quantified by changes in several measures of financial performance for a firm. An Industry Impact Analysis shows financial and competitive impacts on the appliance industry. A Utility Analysis that measures the impacts of the altered energy-consumption patterns on electric utilities. A Environmental Effects analysis, which estimates changes in emissions of carbon dioxide, sulfur oxides, and nitrogen oxides, due to reduced energy consumption in the home and at the power plant. A Regulatory Impact Analysis collects the results of all the analyses into the net benefits and costs from a national perspective. 47 figs., 171 tabs. (JF)

  20. Safety and Security Policy Jobs

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

    Health Safety and Health The DOE Headquarters Safety and Health Program provides information, guidelines, documentation, training, and materials pertaining to many aspects of Safety and Health within the HQ buildings. Question concerning the Headquarters Safety and Health Program can be directed to the Industrial Hygiene and Safety Office on 202-586-1005, or via e-mail to HQSafetyandHealth@hq.doe.gov. Information for Department of Energy Headquarters Personnel The Office of Industrial Hygiene

  1. 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.''

  2. 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...

  3. 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.

  4. 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.

  5. H2 Safety Snapshot - Vol. 2, Issue 2, July 2011 | Department of Energy

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

    2, July 2011 H2 Safety Snapshot - Vol. 2, Issue 2, July 2011 This third issue describes hazard analysis in H2 facility design and operations. PDF icon h2_snapshot_v2i2.pdf More Documents & Publications DOE-HDBK-1100-2004 Safety Planning Guidance for Hydrogen and Fuel Cell Projects Safety Planning Guidance for Hydrogen and Fuel Cell Projects

  6. WIPP Documents - All documents by title

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

    ... Environmental Management System Description DOEWIPP-05-3318 Rev 5 EPA Letter of Approval to Land Dispose of Non-Liquid Polychlorinated Biphenyls (PCBs) at WIPP This document ...

  7. TWRS safety program plan

    SciTech Connect (OSTI)

    Calderon, L.M., Westinghouse Hanford

    1996-08-01

    Management of Nuclear Safety, Industrial Safety, Industrial Hygiene, and Fire Protection programs, functions, and field support resources for Tank Waste Remediation Systems (TWRS) has, until recently, been centralized in TWRS Safety, under the Emergency, Safety, and Quality organization. Industrial hygiene technician services were also provided to support operational needs related to safety basis compliance. Due to WHC decentralization of safety and reengineering efforts in West Tank Farms, staffing and safety responsibilities have been transferred to the facilities. Under the new structure, safety personnel for TWRS are assigned directly to East Tank Farms, West Tank Farms, and a core Safety Group in TWRS Engineering. The Characterization Project Operations (CPO) safety organization will remain in tact as it currently exists. Personnel assigned to East Tank Farms, West Tank Farms, and CPO will perform facility-specific or project-specific duties and provide field implementation of programs. Those assigned to the core group will focus on activities having a TWRS-wide or programmatic focus. Hanford-wide activities will be the responsibility of the Safety Center of Expertise. In order to ensure an effective and consistent safety program for TWRS under the new organization program functions, goals, organizational structure, roles, responsibilities, and path forward must be clearly established. The purpose of the TWRS Safety Program Plan is to define the overall safety program, responsibilities, relationships, and communication linkages for safety personnel under the new structure. In addition, issues associated with reorganization transition are addressed, including training, project ownership, records management, and dissemination of equipment. For the purpose of this document ``TWRS Safety`` refers to all safety professionals and technicians (Industrial Safety, Industrial Hygiene, Fire Protection, and Nuclear Safety) within the TWRS organization, regardless of their location in the organization.

  8. DOE standard: Firearms safety

    SciTech Connect (OSTI)

    1996-02-01

    Information in this document is applicable to all DOE facilities, elements, and contractors engaged in work that requires the use of firearms as provided by law or contract. The standard in this document provides principles and practices for implementing a safe and effective firearms safety program for protective forces and for non-security use of firearms. This document describes acceptable interpretations and methods for meeting Order requirements.

  9. 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...

  10. Nuclear Safety Regulatory Framework

    Energy Savers [EERE]

    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

  11. Safety problems of water-development works designed for land reclamation

    SciTech Connect (OSTI)

    Shchedrin, V. N.; Kosichenko, Yu. M.

    2011-11-15

    A safety declaration is a fundamental document assuring the safety of water-development works, their correspondence to safety criteria, the design, and active technical regulations and rules.

  12. 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).

  13. Document | Department of Energy

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

    Document Document Extracted Pages PDF icon Document More Documents & Publications EIS-0396: Advance Notice of Intent o Prepare an Environmental Impact Statement EIS-0396: Notice of Intent to Prepare a Programmatic Environmental Impact Statement Meeting Minutes: February 20-21, 2007

  14. Worker Safety and Health Rules, Directives, and Technical Standards...

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

    Worker Safety and Health Rules, Directives, and Technical Standards Worker Safety and Health Rules, Directives, and Technical Standards Document Number Title Date 10 CFR 707 ...

  15. Los Alamos National Laboratory (LANL) Unreviewed Safety Question...

    Office of Environmental Management (EM)

    Unreviewed Safety Question (USQ) Process Los Alamos National Laboratory (LANL) Unreviewed Safety Question (USQ) Process The documents included in this listing are additional...

  16. Nuclear Safety Research and Development Annual Report, December...

    Energy Savers [EERE]

    Nuclear Safety Research and Development Annual Report, December 2014 Nuclear Safety Research and Development Annual Report, December 2014 December 8, 2014 This document is the ...

  17. Delegation of Safety Authorities (Jackson - EM) - DOE Directives...

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

    Safety Authorities (Jackson - EM) by jackson DelegationofSafetyAuthorities-JacksonEM.pdf -- PDF Document, 2.38 MB ID: NA Type: Organizations' Assignment of Responsibility...

  18. Integration of Safety Culture Attributes into EFCOG Work Planning...

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

    Planning and Control Guidance Document Integration of Safety Culture Attributes into EFCOG ... Topics Covered: Integration of Safety Culture (SC) Attributes into EFCOG Work Planning and ...

  19. 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

  20. Furnace Standard Analysis Discussion Document

    Office of Environmental Management (EM)

    heating systems could have on overall energy usage, cost, and carbon emissions outcomes. ... Tons 4.5 Metric Tons 3.5 Metric Tons Annual Cost 1,119 1,029 1,806 714 544 ...

  1. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  2. 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'.

  3. 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...

  4. Latest Documents and Notices | Department of Energy

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

    Latest Documents and Notices Latest Documents and Notices RSS May 9, 2016 EIS-0496: Record of Decision San Luis Transmission Project; Alameda, Merced, San Joaquin and Stanislaus Counties, California May 6, 2016 EIS-0525: Draft Programmatic Environmental Impact Statement Nationwide Public Safety Broadband Network Draft Programmatic Environmental Impact Statement for the Eastern United States May 6, 2016 EIS-0525: EPA Notice of Availability of Draft Programmatic Environmental Impact Statement

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

    Broader source: Energy.gov [DOE]

    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.

  6. CRAD, Facility Safety- Nuclear Facility Safety Basis

    Office of Energy Efficiency and Renewable Energy (EERE)

    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 Nuclear Facility Safety Basis.

  7. WIPP RCRA Documents menu

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

    RCRA Documents Menu Permit Modifications Available for Public Comment Final Documents Draft Documents None currently available RCRA-Related Documents EEG Preliminary Comments on the Fundamental Bases of the Characterization Requirements for Disposal of Transuranic Waste at the Waste Isolation Pilot Plant and Findings and Recommendations of the Transuranic Waste Characterization Task Force Final Report EEG Response to NAS WIPP Committee Questions for May 19, 2003 Meeting Quality Assurance Program

  8. Guidance, Reports, and Documents

    Broader source: Energy.gov [DOE]

    The following are annual reports, guidance documents, and regulatory drivers for long-term surveillance and maintenance activities.

  9. TITLE XVII GOVERNING DOCUMENTS

    Broader source: Energy.gov [DOE]

    The following lists documents that provide the statutory and legislative framework for the Title XVII loan guarantee program.

  10. GTT Tranmission Workshop- Documents

    Broader source: Energy.gov [DOE]

    Use the links below to download documents from the GTT Transmission Workshop, held November 1-2, 2012.

  11. Classification Documents and Publications

    Broader source: Energy.gov [DOE]

    Certain documents and publications created or issued by the Office of Classification are available from this page.

  12. 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.

  13. Office of Document Reviews

    Broader source: Energy.gov [DOE]

    The Office of Document Reviews ensures that all documents prepared at DOE Headquarters are properly marked to identify the level and category of protected information they contain (if any) and to ensure that all documents the Department prepares or is required to review under applicable statutes for public release contain no information requiring protection under law, regulations and Executive orders.

  14. Technique for information retrieval using enhanced latent semantic analysis generating rank approximation matrix by factorizing the weighted morpheme-by-document matrix

    DOE Patents [OSTI]

    Chew, Peter A; Bader, Brett W

    2012-10-16

    A technique for information retrieval includes parsing a corpus to identify a number of wordform instances within each document of the corpus. A weighted morpheme-by-document matrix is generated based at least in part on the number of wordform instances within each document of the corpus and based at least in part on a weighting function. The weighted morpheme-by-document matrix separately enumerates instances of stems and affixes. Additionally or alternatively, a term-by-term alignment matrix may be generated based at least in part on the number of wordform instances within each document of the corpus. At least one lower rank approximation matrix is generated by factorizing the weighted morpheme-by-document matrix and/or the term-by-term alignment matrix.

  15. Light-Weight Radioisotope Heater Unit Safety Analysis Report (LWRHU-SAR). Volume I. A. Introduction and executive summary. B. Reference Design Document (RDD)

    SciTech Connect (OSTI)

    Johnson, E.W.

    1985-10-01

    The orbiter and probe portions of the 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 Nuclear Projects (OSNP) 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 addresses the radiological risks which might be encountered by people both at the launch area and worldwide should postulate mission failures or malfunctions occur, which would result 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.

  16. 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.

  17. Document Flowdown Tool

    SciTech Connect (OSTI)

    2010-12-31

    DFTool performs the following: ? A tool to conduct, capture, document, and manage a systematic review and flowdown of contractual requirements/documents to the site, facility, and project-level implementing documents. ? Ability to alert/email document owners to impending changes/updates and impacts to upper-level linked documents. ? Custom ability to generate reports and/or include custom functionalities based on other needs or consideration for site use. ? Ability to be accessed by all personnel (particularly document owners). ? Shows both document flow-up and flow-down. ? Provides visibility where links/flowdown does not currently exist or is incorrect so it can be changed/updated (no visibility currently exists). ? Ability to capture requirements and ultimately link to an existing/proven requirements-based commitment tracking system which can be used (electively) at the facility/organizational-level

  18. Document Flowdown Tool

    Energy Science and Technology Software Center (OSTI)

    2010-12-31

    DFTool performs the following: • A tool to conduct, capture, document, and manage a systematic review and flowdown of contractual requirements/documents to the site, facility, and project-level implementing documents. • Ability to alert/email document owners to impending changes/updates and impacts to upper-level linked documents. • Custom ability to generate reports and/or include custom functionalities based on other needs or consideration for site use. • Ability to be accessed by all personnel (particularly document owners). •more » Shows both document flow-up and flow-down. • Provides visibility where links/flowdown does not currently exist or is incorrect so it can be changed/updated (no visibility currently exists). • Ability to capture requirements and ultimately link to an existing/proven requirements-based commitment tracking system which can be used (electively) at the facility/organizational-level« less

  19. Technical support document: Energy efficiency standards for consumer products: Refrigerators, refrigerator-freezers, and freezers including draft environmental assessment, regulatory impact analysis

    SciTech Connect (OSTI)

    1995-07-01

    The Energy Policy and Conservation Act (P.L. 94-163), as amended by the National Appliance Energy Conservation Act of 1987 (P.L. 100-12) and by the National Appliance Energy Conservation Amendments of 1988 (P.L. 100-357), and by the Energy Policy Act of 1992 (P.L. 102-486), provides energy conservation standards for 12 of the 13 types of consumer products` covered by the Act, and authorizes the Secretary of Energy to prescribe amended or new energy standards for each type (or class) of covered product. The assessment of the proposed standards for refrigerators, refrigerator-freezers, and freezers presented in this document is designed to evaluate their economic impacts according to the criteria in the Act. It includes an engineering analysis of the cost and performance of design options to improve the efficiency of the products; forecasts of the number and average efficiency of products sold, the amount of energy the products will consume, and their prices and operating expenses; a determination of change in investment, revenues, and costs to manufacturers of the products; a calculation of the costs and benefits to consumers, electric utilities, and the nation as a whole; and an assessment of the environmental impacts of the proposed standards.

  20. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    L.L. Swanson

    2005-04-06

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of procedures that will be used to operate the system.

  1. CRAD, NNSA- Nuclear Explosive Safety (NES)

    Broader source: Energy.gov [DOE]

    CRAD for Nuclear Explosive Safety (NES). Criteria Review and Approach Documents (CRADs) that can be used to conduct a well-organized and thorough assessment of elements of safety and health programs.

  2. Hydrogen Safety Panel | Department of Energy

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

    7_weiner.pdf More Documents & Publications Hydrogen Safety: First Responder Education Hydrogen Safety Knowledge Tools NanoCapillary Network Proton Conducting Membranes for High Temperature Hydrogen/Air Fuel Cells

  3. A safety overview of Sandia National Laboratories' reactor facilities

    SciTech Connect (OSTI)

    Philbin, J.S.

    1989-04-01

    This report provides an overview of Sandia National Laboratories' safety policies and practices supporting the operation of Sandia's nuclear reactor facilities. These policies and practices have evolved from Sandia's 30 years of experience and leadership in the design, construction, and operation of steady-state and pulse research reactors. The report illustrates how Sandia has implemented DOE orders and research reactor standards with the goal of reducing risks to the lowest reasonable levels for its employees, contractors, the public, and the environment. The impact of DOE orders and standards on virtually all aspects of reactor operations and administration is illustrated. Included in the report are descriptions of safety documentation (Technical Specifications and Safety Analysis Reports); the facility safety review system for addressing radiological protection and other environmental, safety and health issues; experiment activities; quality assurance; training and certification; and emergency planning.

  4. Integration of Environment, Safety, and Health into Facility Disposition Activities

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

    1998-05-01

    Volume One of this Standard has been revised to provide a Department of Energy (DOE) approved methodology for preparing a Documented Safety Analysis (DSA) for decommissioning of nuclear facilities, as well as environmental restoration activities that involve work not done within a permanent structure. Methodologies provided in this Standard are intended to be compliant with Title 10 of the Code of Federal Regulations (CFR) Part 830, Nuclear Safety Management, Subpart B, Safety Basis Requirements. Volume Two contains the appendices that provide additional environment, safety and health (ES&H) information to complement Volume 1 of this Standard. Volume 2 of the Standard is much broader in scope than Volume 1 and satisfies several purposes. Integrated safety management expectations are provided in accordance with facility disposition requirements contained in DOE O 430.1B, Real Property Asset Management.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Electrical Safety Manual | The Ames Laboratory

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

    Electrical Safety Manual Version Number: 0 Document Number: Manual 10200.007 Effective Date: 032012 File (public): PDF icon manual10200.007rev0.pdf...

  10. Bylaws and Other Documents

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

    Bylaws and Other Documents Postdoc Association Bylaws Bylaws and information about rules of order, meeting conduct, and motions. Contact Postdoc Program Office Email How to go about properly conducting business at a General Assembly Meeting Refer to the postdoc bylaws and other documents, below, for information on how to go about properly conducting business at a LAPA General Assembly (GA) Meeting. The first two documents will be available on the tables at the GA meetings for reference: Bylaws

  11. CAPTURE DOCUMENT ORAUTEAM

    Office of Legacy Management (LM)

    DATA CAPTURE DOCUMENT ORAUTEAM ---- Dose Reconstruction ~v~:7 DISCOVERY AND REVIEW dA'~ Project for NIOSH The attached document may contain Privacy Act data. This information is protected by the Privacy Act, 5 U.S.C. §552a; disclosure to any third party without written consent of the individual to whom the information pertains is strictly prohibited. Data Capture Team or Other ORAU Team Member Capturing Data: Complete all information that applies to the data/document being submitted lor

  12. Community impact documents

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

    Community impact documents Community impact documents Fact sheets, program summaries, and other documents provide insight into the Laboratory's community efforts and impact in Northern New Mexico. Contacts Kathy Keith Community Relations & Partnerships (505) 665-4400 Email Making a difference in Northern New Mexico Fact sheets by county Los Alamos (pdf) Rio Arriba (pdf) San Miguel and Mora (pdf) Santa Fe (pdf) Taos (pdf) General Community Commitment Plan (pdf) | Archive Community Leaders

  13. 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.

  14. 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.

  15. Documents for Foreign Nationals

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

    present the following documents when they register at the ALS User Services Office: Passport: Passport must show issue and expiration date, passport number, and photo. I-94...

  16. ARM - Program Documents

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

    Videos Publication Resources Submit a Publication Publishing Procedures ARM Style Guide (PDF, 448KB) Acronyms Glossary Logos Contacts RSS for Publications Program Documents Annual...

  17. Hazard Baseline Documentation

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

    1995-12-04

    This standard establishes uniform Office of Environmental Management (EM) guidance on hazard baseline documents that identify and control radiological and non-radiological hazards for all EM facilities.

  18. SPEAR3 | Technical Documentation

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

    Technical Documentation Hardware Component Database: Mechanical Electrical Drawing shor tcuts BPM Development SSRL | SLAC | Stanford University | SSRL Computing | SLAC Computing last updated: NOV 25, 2003 a.mueller

  19. Archived Publications and Documents

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

    Archived Publications and Documents Print ALS 20-Year Roadmap, 2003 Presentations made to the BESAC subcommittee on the 20-Year BES Facilities Roadmap, with references and relevant...

  20. Plans, Updates, Regulatory Documents

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

    Permit for Storm Water Documents Individual Permit NPDES No. NM0030759 (pdf) Storm Water Plans Site Discharge Pollution Prevention Plan (SDPPP) Volume 1 - Los AlamosPueblo...

  1. Documents | Department of Energy

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

    Documents Memorandum from Secretary Moniz on the Freedom of Information Act (PDF) Agency Improvement Plan (PDF) FOIA Backlog Reduction Goals for Fiscal Years 2008, 2009, and 2010 ...

  2. 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 show that these rates and maximum concentrations are so low as to make it unlikely that the LFL concentration would ever be approached. The second objective of this analysis is to review the data provided by two separate flammable gas measurement systems on each tank being saltwell pumped to see if there is an unnecessary redundancy. Eliminating redundant measurement systems would provide cost savings if the quality of data and resultant margin of safety during saltwell pumping activity are not compromised.

  3. Enron Documents 2000 | Department of Energy

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

    0 Enron Documents 2000 Enron Documents for 2000, enron2000.pdf PDF icon Enron Documents 2000 More Documents & Publications Enron Documents 1996 FOIA - Enron Documents Documents ...

  4. Enron Documents 1998 | Department of Energy

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

    8 Enron Documents 1998 Enron Documents 1998 PDF icon Enron Documents 1998 More Documents & Publications FOIA - Enron Documents Enron Documents 1993 Enron Documents 1996

  5. Enron Documents 1993 | Department of Energy

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

    3 Enron Documents 1993 Enron Documents 1993. enron1993.pdf PDF icon Enron Documents 1993 More Documents & Publications FOIA - Enron Documents Enron Documents 1998 Enron Documents ...

  6. Enron Documents 1996 | Department of Energy

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

    6 Enron Documents 1996 Enron Documents 1996 PDF icon Enron Documents 1996 More Documents & Publications Enron Documents 2000 FOIA - Enron Documents Enron Documents 2001

  7. PAT-1 safety analysis report addendum author responses to request for additional information.

    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 National Nuclear Security Administration (NNSA) submitted SAND Report SAND2009-5822 to NRC that documented the incorporation of plutonium (Pu) metal as a new payload for the PAT-1 package. NRC responded with a Request for Additional Information (RAI), identifying information needed in connection with its review of the application. The purpose of this SAND report is to provide the authors responses to each RAI. SAND Report SAND2010-6106 containing the proposed changes to the Addendum is provided separately.

  8. Quality Procedure- Document Review

    Broader source: Energy.gov [DOE]

    This procedure establishes the process for the reivew and approval of Environmental Managemnt (EM) Quality Assurance (QA) program documents as performed by the Office of Standards and Quality Assurance. This procedure also establishes the process for review and approval of other EM documents external to the EM Headquarters Office of Standards and Quality Assurance.

  9. IDC System Specification Document.

    SciTech Connect (OSTI)

    Clifford, David J.

    2014-12-01

    This document contains the system specifications derived to satisfy the system requirements found in the IDC System Requirements Document for the IDC Reengineering Phase 2 project. Revisions Version Date Author/Team Revision Description Authorized by V1.0 12/2014 IDC Reengineering Project Team Initial delivery M. Harris

  10. Cray XT Documentation

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

    Cray XT Documentation Cray XT Documentation </p> Last edited: 2011-04-06 11:35:53

  11. Appendix A: Conceptual Documents

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

    NREL-Research Support Facilities TABLE OF CONTENTS Solicitation No. RFJ-8-77550 Appendix A: Conceptual Documents Page 1 of 299 February 6, 2008 Appendix A CONCEPTUAL DOCUMENTS TABLE OF CONTENTS INTRODUCTION ...................................................................................................................... 2 PART 1-PROCEDURES .......................................................................................................... 3 PART 2-PROGRAM

  12. Technical Safety Requirements for the Waste Storage Facilities May 2014

    SciTech Connect (OSTI)

    Laycak, D. T.

    2014-04-16

    This document contains the Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Building 693 (B693) Yard Area of the Decontamination and Waste Treatment Facility (DWTF) at LLNL. The TSRs constitute requirements for safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analyses for the Waste Storage Facilities (DSA) (LLNL 2011). The analysis presented therein concluded that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs.

  13. 2015 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 2010 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.

  4. 2011 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.

  5. 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 reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 10{sup 6} cm{sup −1}) the limits were exceeded.

  6. Structure-Soil-Structure Interaction Effects: Seismic Analysis of Safety-Related Collocated Structures

    Office of Environmental Management (EM)

    STRUCTURE-SOIL- STRUCTURE INTERACTION AT SRS Structural Mechanics - SRS October 25, 2011 1 Objective Determination of Structure Soil Structure Interaction (SSSI) effects, if any between large and more massive Process Building (PB) and Exhaust Fan Building (EFB). Results of the SSSI analysis were compared with those from Soil Structure Interaction (SSI) analysis of the individual buildings, for the following parameters: * In-structure floor response spectra (ISRS) * Transfer functions * Relative

  7. Sandia Energy - Reference Model Documents

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

    Documents Home Stationary Power Energy Conversion Efficiency Water Power Reference Model Project (RMP) Reference Model Documents Reference Model DocumentsTara Camacho-Lopez2015-05-...

  8. 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.

  9. 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.

  10. Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants

    SciTech Connect (OSTI)

    Kisner, Roger A; Mullens, James Allen; Wilson, Thomas L; Wood, Richard Thomas; Korsah, Kofi; Qualls, A L; Muhlheim, Michael David; Holcomb, David Eugene; Loebl, Andy

    2007-08-01

    Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

  11. 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.

  12. Microsoft Word - Document2

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

    A Village of Innovation Partners If you read our last newsletter, you know that we have ... health, safety, and durability for lower cost of ownership (e.g., utility savings that ...

  13. 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 &...

  14. Microsoft Word - Document2

    Buildings Energy Data Book [EERE]

    09 D I S C L A I M E R This document was designed for the internal use of the United States Department of Energy. This document will be occasionally updated and, therefore, this copy may not reflect the most current version. This document was prepared as 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

  15. DOCUMENT RELEASE FORM C

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

    RELEASE FORM C (1) Document Number: RPP-RPT-431 74 (2) Revision Number: (3) Effective Date: 9/30/2009 (4) Document Type: El Digital Image El Hard copy (a) Number of pages (including the DRF) or 107 E PDF E] Video number of digital images (5) Release Type 0 New El Cancel El Page Change El complete Revision (6) Document Title: 2009 Auto-TOR for Tank 241 -T-204 (7) Change/Release Initial Issuance Description: (8) Change Initial Issuance Justification: (9) Associated (a) Structure Location: (c)

  16. 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

  17. The use of representative cases in hazard analysis of the tank waste remediation system at Hanford. The information in this document is a combination of HNF-SA-3168-A {ampersand} HNF-SA-3169-A - The control identification process

    SciTech Connect (OSTI)

    Niemi, B.J.

    1997-04-24

    During calendar year 1996, Duke Engineering and Services Hanford, Inc. conducted a safety analysis in accordance with DOE-STD-3009-94 as part of the development of a Final Safety Analysis Report (TSAR) for the Tank Waste Remediation System (TWRS) at the DOE Hanford site. The scope of the safety analysis of TWRS primarily addressed 177 large underground liquid waste storage tanks and associated equipment for transferring waste to and from tanks. The waste in the tanks was generated by the nuclear production and processing facilities at Hanford. The challenge facing the safety analysis team was to efficiently analyze the system within the time and budget allotted to provide the necessary and sufficient information for accident selection, control identification, and justification on the acceptability of the level of safety of TWRS. It was clear from the start that a hazard and accident analysis for each of the 177 similar tanks and supporting equipment was not practical nor necessary. For example, many of the tanks were similar enough that the results of the analysis of one tank would apply to many tanks. This required the development and use of a tool called the ''Hazard Topography''. The use of the Hazard Topography assured that all tank operations and configurations were adequately assessed in the hazard analysis and that the results (e.g., hazard identification and control decisions) were appropriately applied to all tanks and associated systems. The TWRS Hazard Topography was a data base of all the TWRS facilities (e.g., tanks, diversion boxes, transfer lines, and related facilities) along with data on their configuration, material at risk (MAR), hazards, and known safety related phenomenological issues. Facilities were then classified into groups based on similar combinations of configuration, MAR, hazards and phenomena. A hazard evaluation was performed for a tank or facility in each group. The results of these evaluations, also contained in a data base, were then mapped back to all TWRS facilities and used to select candidate accidents for the SAR. The Hazard Topography and hazard evaluation results were then used to support the identification of controls that address all TWRS facilities.

  18. WIPP Documents - All documents by number

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

    ... DOEEIS 0026-S-2 WIPP Disposal Phase Supplemental Environmental Impact Statement (SEIS-II) This ... of monitoring and measurements, procedures for lab analysis, QA requirements, ...

  19. Safety and Health | Department of Energy

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

    Health Safety and Health The DOE Headquarters Safety and Health Program provides information, guidelines, documentation, training, and materials pertaining to many aspects of Safety and Health within the HQ buildings. Question concerning the Headquarters Safety and Health Program can be directed to the Industrial Hygiene and Safety Office on 202-586-1005, or via e-mail to HQSafetyandHealth@hq.doe.gov. Information for Department of Energy Headquarters Personnel The Office of Industrial Hygiene

  20. Nuclear Safety Regulatory Framework | Department of Energy

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

    Nuclear Safety Regulatory Framework Nuclear Safety Regulatory Framework February 2012 Presentation that outlines the rules, policies and orders that comprise the Department of Energy Nuclear Safety Regulatory Framework. PDF icon Nuclear Safety Regulatory Framework More Documents & Publications Summary Pamphlet, Nuclear Safety at the Department of Energy Notice of Violation, UChicago Argonne, LLC - WEA-2009-04 Independent Oversight Assessment, Waste Treatment and Immobilization Plant -

  1. Quality Procedure- Document Control

    Broader source: Energy.gov [DOE]

    This procedure establishes the responsibilities and process for preparing, managing, and revising EM Headquarters Office of Standards and Quality Assurance controlled documents in accordance with EM-QA-001, Environmental Management Quality Assurance Program.

  2. Geochemistry Technical Basis Document

    SciTech Connect (OSTI)

    Benedict, Jr, F Christopher; Rose, Timothy P; Thomas, James M; Waddell, Richard; Jacobson, Roger

    2004-03-18

    This document presents a methodology whereby geochemical data can more effectively contribute to the development , calibration, and verification of groundwater flow and slute transport models for the Underground Test Area (UGTA) Project.

  3. WIPP Documents - EPA Certification

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

    EPA Letter of Approval to Land Dispose of Non-Liquid Polychlorinated Biphenyls (PCBs) at WIPP This document provides EPA approval for the disposal of TRU and TRU-Mixed waste ...

  4. QER Document Library

    Broader source: Energy.gov [DOE]

    EPSA prepared numerous reports and technical workshops as the QER was developed. These documents will be posted here in the days ahead. Check back for new and updated materials.

  5. Document (17k)

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

    document.xml.rels &3; wordrelssettings.xml.rels &3; wordrels &3; worddocument.xml SAM PLE APPROVAL MEMORANDUM FOR ADMINISTRATIVE CHANGES ONLY (Updated 2-26-2013) (NOTE: Per...

  6. Controlled Document Tracking Software

    Energy Science and Technology Software Center (OSTI)

    1992-08-24

    MANTRACK is an automated, controlled document tracking system which does the following and reduces staff time required to perform these tasks: generates transmittal letters/receipts for every controlled copy issued (merged from a current distribution list), tracks the return of transmittal receipts, facilitates the check-in of the large number of transmittal receipts returned (using a barcode reader), generates a reminder list which prompts the cyclic review and evaluation of existing documents, generates overdue reminders for themore » return of past-due transmittal receipts, tracks the number of Procedure Change Directives (PCD) currently in effect for each procedure, generates and maintains current distribution lists for each document, generates a current table of contents when updates to the document (usually a procedure manual) are made.« less

  7. 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)

  8. 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.

  9. DOE HQ Occupational Safety and Health Program | Department of Energy

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

    Occupational Safety and Health Program DOE HQ Occupational Safety and Health Program HQ Occupational Safety and Health Program Procedures PDF icon DOE HQ Occupational Safety and Health Program More Documents & Publications HQ Confined Space Program, Policy 2010-001 Independent Oversight Review, Department of Energy Contractor - August 2000 FAQS Qualification Card - Occupational Safety

  10. Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 |

    Energy Savers [EERE]

    Department of Energy Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 September 2012 Quickstart guide on how to use the features of Nuclear Safety Information dashboard tool. PDF icon Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 More Documents & Publications Development of the Nuclear Safety Information Dashboard - September 2012 Responses to Frequently Asked

  11. The System 80+ Standard Plant design control document. Volume 17

    SciTech Connect (OSTI)

    1997-12-31

    This Design Control Document (DCD) is a repository of information comprising the System 80+{trademark} Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ``Tier 1``] and the Approved Design Material (ADM) [i.e., ``Tier 2``] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 2-7 and appendix 15A for section 15 (Accident Analysis) of the ADM Design and Analysis. Topics covered in these parts are: decrease in heat removal; decrease in RCS flow rate; power distribution anomalies; increase in RCS inventory; decrease in RCS inventory; release of radioactive materials. The appendix covers radiological release models. Also contained here are five technical specifications for section 16 (Technical Specifications) of the ADM Design and Analysis. They are: TS 1.0 Use and Applications; TS 2.0 Safety Limits; TS 3.0 LCO Availability; TS 3.1 Reactivity Control; and TS 3.2 Power Distribution.

  12. 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.

  13. 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...

  14. 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...

  15. 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 available to provide improved understanding of the VHTR during accidents.

  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. 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.

  18. High level waste storage tanks 242-A evaporator standards/requirement identification document

    SciTech Connect (OSTI)

    Biebesheimer, E.

    1996-01-01

    This document, the Standards/Requirements Identification Document (S/RIDS) for the subject facility, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment. It lists those source documents from which requirements were extracted, and those requirements documents considered, but from which no requirements where taken. Documents considered as source documents included State and Federal Regulations, DOE Orders, and DOE Standards

  19. 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.

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

    SciTech Connect (OSTI)

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in 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 compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

  1. 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.

  2. 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).

  3. DOCUMENT RELEASE FORM

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

    8 (2) Revision Number: 0 (3) Effective Date: 9/28/2009 (4) Document Type: [] Digital Image El Hard copy (a) Number of pages (including the DRF) or 74 0PDF E] Video number of digital images (5) Release Type E New El cancel l Page Change El complete Revision (6) Document Title: 2009 Auto-TCR for Tank 241-13-204 (7) Change/Release Initial Issuance Description: (8) Change Initial Issuance Justification: (9) Associated (a) Structure Location: (c) Building Number: Structure, System, and Component N/A

  4. DOCUMENT RELEASE FORM

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

    31 97 (2) Revision Number: 0 -- (3) Effective Date: 9/30/2009 (4) Document Type: E] Digital Image l Hard copy (a) Number of pages (including the DRF) or 84 0 POE E Video number of digital images (5) Release Type Z New 1:1 Cancel liiPage Change Elcomplete Revision (6) Document Title: 2009 Auto-TOR for Tank 241 -TY-1 05 (7) Change/Release Initial Issuance Description: (8) Change Initial Issuance Justification: (9) Associated (a) Structure Location: (c) Building Number: Structure, System, and

  5. IXPUG ISC15 Documents

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

    Training » IXPUG ISC15 Documents IXPUG ISC15 Documents Sort by: Default | Name | Date (low-high) | Date (high-low) | Source | Category BoF: IXPUG ISC15 BoF Introduction July 15, 2015 | Author(s): Thomas Steinke, ZIB | Download File: 1Welcome.pdf | pdf | 98 KB Workshop: Intel® Xeon Phi(tm) Processor "Knights Landing" Architectural Overview Author(s): Avinash Sodani, Senior Principal Engineer, Intel Corporation Chief Architect, Knights Landing Processor | Download File:

  6. Documents for Foreign Nationals

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

    Documents for Foreign Nationals Print All foreign nationals (non-U.S. citizens) intending to work at the ALS must ensure that they have all relevant travel and visa documents completed and approved before they will be permitted to work at Berkeley Lab. Owing to increased security measures, it is particularly important to begin the visa application process early in order to avoid delays that may result in lost beamtime, etc. NOTE: Researchers who are citizens of, or were born in, T4 countries

  7. CRAD, Explosives Safety- February 19, 2015 (EA CRAD 32-01, Rev. 0)

    Broader source: Energy.gov [DOE]

    CRAD, Explosives Safety – February 19, 2015 (EA CRAD 32-01, Rev. 0) Explosives Safety Criteria Review and Approach Document (EA CRAD 32-01, Rev. 0)

  8. Document clustering methods, document cluster label disambiguation methods, document clustering apparatuses, and articles of manufacture

    DOE Patents [OSTI]

    Sanfilippo, Antonio; Calapristi, Augustin J.; Crow, Vernon L.; Hetzler, Elizabeth G.; Turner, Alan E.

    2009-12-22

    Document clustering methods, document cluster label disambiguation methods, document clustering apparatuses, and articles of manufacture are described. In one aspect, a document clustering method includes providing a document set comprising a plurality of documents, providing a cluster comprising a subset of the documents of the document set, using a plurality of terms of the documents, providing a cluster label indicative of subject matter content of the documents of the cluster, wherein the cluster label comprises a plurality of word senses, and selecting one of the word senses of the cluster label.

  9. Electrical Safety Occurrences | Department of Energy

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

    Electrical Safety Occurrences Electrical Safety Occurrences June 26, 2014 Monthly Analysis of Electrical Safety Occurrences - April 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. October 29, 2013 Monthly Analysis of Electrical Safety Occurrences - September 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by

  10. 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.

  11. Documents | Department of Energy

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

    Documents Memorandum from Secretary Moniz on the Freedom of Information Act (PDF) Agency Improvement Plan (PDF) FOIA Backlog Reduction Goals for Fiscal Years 2008, 2009, and 2010 (PDF) Ultra-Deepwater and Unconventional Natural Gas and Other Petroleum Resources (COMING SOON) Updated Report on Executive Order 13392 Implementation (PDF) Certification DOE Reading Rooms (pdf)

  12. Program Documents | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Program Documents Reference Guide Process

  13. Safety and Security Interface Technology Initiative

    SciTech Connect (OSTI)

    Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

    2007-05-01

    Safety and Security Interface Technology Initiative Mr. Kevin J. Carroll Dr. Robert Lowrie, Dr. Micheal Lehto BWXT Y12 NSC Oak Ridge, TN 37831 865-576-2289/865-241-2772 carrollkj@y12.doe.gov Work Objective. Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. Supporting Excellence in Operations Through Safety Analysis, (workshop theme) includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is Safeguards/Security Integration with Safety. This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis/Security Documentation Integration, Configuration Control, and development of a shared tool box of information/successes. Specific Benefits. The expectation or end state resulting from the topical report and associated implementation plan includes: (1) A recommended process for handling the documentation of the security and safety disciplines, including an appropriate change control process and participation by all stakeholders. (2) A means to package security systems with sufficient information to help expedite the flow of that system through the process. In addition, a means to share successes among sites, to include information and safety basis to the extent such information is transportable. (3) Identification of key security systems and associated essential security elements being installed and an arrangement for the sites installing these systems to host an appropriate team to review a specific system and determine what information is exportable. (4) Identification of the security systems essential elements and appropriate controls required for testing of these essential elements in the facility. (5) The ability to help refine and improve an agreed to control set at the manufacture stage.

  14. ARM - ARM Safety Policy

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

    Safety Policy About Become a User Recovery Act Mission FAQ Outreach Displays History Organization Participants Facility Statistics Forms Contacts Facility Documents ARM Management Plan (PDF, 1.3MB) Field Campaign Guidelines (PDF, 574KB) ARM Climate Research Facility Expansion Workshop (PDF, 1.46MB) Facility Activities ARM and the Recovery Act Contributions to International Polar Year Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send ARM Safety

  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. Mixed waste characterization reference document

    SciTech Connect (OSTI)

    1997-09-01

    Waste characterization and monitoring are major activities in the management of waste from generation through storage and treatment to disposal. Adequate waste characterization is necessary to ensure safe storage, selection of appropriate and effective treatment, and adherence to disposal standards. For some wastes characterization objectives can be difficult and costly to achieve. The purpose of this document is to evaluate costs of characterizing one such waste type, mixed (hazardous and radioactive) waste. For the purpose of this document, waste characterization includes treatment system monitoring, where monitoring is a supplement or substitute for waste characterization. This document establishes a cost baseline for mixed waste characterization and treatment system monitoring requirements from which to evaluate alternatives. The cost baseline established as part of this work includes costs for a thermal treatment technology (i.e., a rotary kiln incinerator), a nonthermal treatment process (i.e., waste sorting, macronencapsulation, and catalytic wet oxidation), and no treatment (i.e., disposal of waste at the Waste Isolation Pilot Plant (WIPP)). The analysis of improvement over the baseline includes assessment of promising areas for technology development in front-end waste characterization, process equipment, off gas controls, and monitoring. Based on this assessment, an ideal characterization and monitoring configuration is described that minimizes costs and optimizes resources required for waste characterization.

  17. Los Alamos Field Office NEPA Documents | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Los Alamos Field Office NEPA Documents NEPA Documents Below are links that provide NEPA documents which govern the Los Alamos National Laboratory operational envelope. This website is organized into three categories: Environmental Impact Statements (EIS), which includes Supplemental Analysis (SA); Environmental Assessments (EA) and Categorical Exclusions (CX). Under each of these sections, you will find a list of documents organized by document number, which is linked to an

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

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

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

  19. Document Imaging | Department of Energy

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

    Document Imaging Document Imaging Converting Paper Documents into Electronic Files Converting paper documents into electronic files helps us manage, store, access and archive the organizational information we have "locked up" in paper documents. Utilizing high-quality document scanners, a top-end six-engine Optical Character Recognition (OCR) system and maintaining Quality Controls to provide a successful Imaging solution. Once converted, these electronic files can be indexed and

  20. NEPA Documents | Department of Energy

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

    Services » NEPA Documents NEPA Documents To see a list of available NEPA documents by type and sorted by publication date, click on the links below. Documents listed in this section are available to members of the public. To see documents not electronically available to the public see the Secure NEPA Documents page for more information. Categorical Exclusion (CX) Determinations Categorical exclusions are categories of actions that DOE has determined, by regulation, do not individually or