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Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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1

Manpower analysis in transportation safety. Final report  

DOE Green Energy (OSTI)

The project described provides a manpower review of national, state and local needs for safety skills, and projects future manning levels for transportation safety personnel in both the public and private sectors. Survey information revealed that there are currently approximately 121,000 persons employed directly in transportation safety occupations within the air carrier, highway and traffic safety, motor carrier, pipeline, rail carrier, and marine carrier transportation industry groups. The projected need for 1980 is over 145,000 of which over 80 percent will be in highway safety. An analysis of transportation tasks is included, and shows ten general categories about which the majority of safety activities are focused. A skills analysis shows a generally high level of educational background and several years of experience are required for most transportation safety jobs. An overall review of safety programs in the transportation industry is included, together with chapters on the individual transportation modes.

Bauer, C.S.; Bowden, H.M.; Colford, C.A.; DeFilipps, P.J.; Dennis, J.D.; Ehlert, A.K.; Popkin, H.A.; Schrader, G.F.; Smith, Q.N.

1977-05-01T23:59:59.000Z

2

Rankine bottoming cycle safety analysis. Final report  

SciTech Connect

Vector Engineering Inc. conducted a safety and hazards analysis of three Rankine Bottoming Cycle Systems in public utility applications: a Thermo Electron system using Fluorinal-85 (a mixture of 85 mole % trifluoroethanol and 15 mole % water) as the working fluid; a Sundstrand system using toluene as the working fluid; and a Mechanical Technology system using steam and Freon-II as the working fluids. The properties of the working fluids considered are flammability, toxicity, and degradation, and the risks to both plant workers and the community at large are analyzed.

Lewandowski, G.A.

1980-02-01T23:59:59.000Z

3

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

Science Conference Proceedings (OSTI)

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.

DAUTEL, W.A.

2002-01-10T23:59:59.000Z

4

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

Science Conference Proceedings (OSTI)

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.

Alan Bond

2001-04-01T23:59:59.000Z

5

Hanford Sludge Treatment Project 105-KW Final Safety Analysis Report Review, August 2011  

NLE Websites -- All DOE Office Websites (Extended Search)

Site Visit Report Site Visit Report Sludge Treatment Project 105-KW Final Safety Analysis Report Review May 2011 August 2011 Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Introduction ............................................................................................................................................ 1 2.0 Background ............................................................................................................................................ 1 3.0 Scope ...................................................................................................................................................... 1 4.0 Results .................................................................................................................................................... 2

6

Hanford Sludge Treatment Project 105-KW Final Safety Analysis Report Review, August 2011  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Site Visit Report Site Visit Report Sludge Treatment Project 105-KW Final Safety Analysis Report Review May 2011 August 2011 Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Introduction ............................................................................................................................................ 1 2.0 Background ............................................................................................................................................ 1 3.0 Scope ...................................................................................................................................................... 1 4.0 Results .................................................................................................................................................... 2

7

CFAST Computer Code Application Guidance for Documented Safety Analysis, Final Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Final CFAST Code Guidance Final CFAST Code Guidance CFAST Computer Code Application Guidance for Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 July 2004 DOE/NNSA-DP Technical Report CFAST Computer Code Application Guidance Final Report July 2004 ii INTENTIONALLY BLANK. DOE/NNSA-DP Technical Report CFAST Computer Code Application Guidance Final Report July 2004 iii FOREWORD This document provides guidance to Department of Energy (DOE) facility analysts in the use of the CFAST computer software for supporting Documented Safety Analysis applications. Information is provided herein that supplements information found in the CFAST documentation

8

FINAL SAFETY ANALYSIS REPORT. SNAP III THERMOELECTRIC GENERATOR  

SciTech Connect

The SNAP-III thermoelectric generator procedures power from the decay heat of 2100 curies of Po/sup 210/. This generator is to be used as a source of auxiliary power in a terrestrial satellite. For purposes of analysis, the satellite system postulated is launched from the Pacific Missile Bange into a 275- statute mile polar orbit with an orbital lifetime of about 1 year. Po/sup 210/ is an alpha emitter having a half life of 138 days and alpha and gamma decay energies of 5.3 and 0.8 mev, respectively. It is a natural component of the earth's crust, as a member of the uranium disintegration series. Sampling of polonium in the biosphere was conducted specifically for this program to determine background radiation levels. Since the fuel is primarily an alpha emitter, there is no direct radiation problem. An analysis was performed to determine the ability of the fuel container to withstand the various thermal, mechanical, and chemical forces imposed upon the generator by vehicle failures. Where theoretical analysis was impossible and experimental evidence was desired, capsules and generators were tested under simulated missile-failure conditions, Thus, the safety limits of SNAP-III in a satellite application were defined. SNAP-III is designed to be aerothermodynamically consumed on reentry into the earth's atmosphere so that the polonium will be dispersed as aerosols in the upper stratosphere. Since heating rates will be lower for aborts occurring prior to orbiting, 65 abort cases have been considered to define the general consequences of vehicle failures. The spatial and temporal relations of vehicle aborts are summarized in cartographic and tabular form. (auth)

Hagis, W.; Dix, G.P.

1960-06-01T23:59:59.000Z

9

Hanford Sludge Treatment Project 105-KW Final Safety Analysis...  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis Report (HNF-SD-WM-SAR-062, Revision 14C) for the Sludge Treatment Project at the Hanford Site. 2.0 BACKGROUND The Sludge Treatment Project manages the removal of...

10

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

Science Conference Proceedings (OSTI)

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.

Not Available

1988-12-15T23:59:59.000Z

11

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

SciTech Connect

The Accident Model Document (AMD) is the second volume of the three volume Final Safety Analysis Report (FSAR) for the Galileo outer planetary space science mission. This mission employs Radioisotope Thermoelectric Generators (RTGs) as the prime electrical power sources for the spacecraft. Galileo will be launched into Earth orbit using the Space Shuttle and will use the Inertial Upper Stage (IUS) booster to place the spacecraft into an Earth escape trajectory. The RTG's employ silicon-germanium thermoelectric couples to produce electricity from the heat energy that results from the decay of the radioisotope fuel, Plutonium-238, used in the RTG heat source. The heat source configuration used in the RTG's is termed General Purpose Heat Source (GPHS), and the RTG's are designated GPHS-RTGs. The use of radioactive material in these missions necessitates evaluations of the radiological risks that may be encountered by launch complex personnel as well as by the Earth's general population resulting from postulated malfunctions or failures occurring in the mission operations. The FSAR presents the results of a rigorous safety assessment, including substantial analyses and testing, of the launch and deployment of the RTGs for the Galileo mission. This AMD is a summary of the potential accident and failure sequences which might result in fuel release, the analysis and testing methods employed, and the predicted source terms. Each source term consists of a quantity of fuel released, the location of release and the physical characteristics of the fuel released. Each source term has an associated probability of occurrence. 27 figs., 11 tabs.

Not Available

1988-12-15T23:59:59.000Z

12

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

Science Conference Proceedings (OSTI)

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.

NONE

1994-10-01T23:59:59.000Z

13

FINAL SAFETY ANALYSIS REPORT--SNAP 1A RADIOISOTOPE FUELED THERMOELECTRIC GENERATOR  

SciTech Connect

The safety aspects involved in utilizing the Task 2 radioisotope-powered thermoelectric generator in a terrestrial satellite are described. It is based upon a generalized satellite mission having a 600-day orbital lifetime. A description of the basic design of the generator is presented in order to establish the analytical model. This includes the generator design, radiocerium fuel properties, and the fuel core. The transport of the generator to the launch site is examined, including the shipping cask, shipping procedures, and shipping hazards. A description of ground handling and vehicle integration is presented including preparation for fuel transfer, transfer, mating of generators to final stage, mating final stage to booster, and auxiliary support equipment. The flight vehicle is presented to complete the analytical model. Contained in this chapter are descriptions of the booster-sustainer, final stage, propellants, and built-in safety systems. The typical missile range is examined with respect to the launch complex and range safety characteristics. The shielding of the fuel is discussed and includes both dose rates and shield thicknesses required. The bare core, shielded generator, fuel transfer operation and dose rates for accidental conditions are treated. mechanism of re-entry from the successful mission is covered. Radiocerium inventories with respect to time and the chronology of re-entry are specifically treated. The multiplicity of conditions for aborted missions is set forth. The definition of aborted missions is treated first in order to present the initial conditions. Following this, a definition of the forces imposed upon the generator is presented. The aborted missions is presented. A large number of initial vehicle failure cases is narrowed down into categories of consequences. Since stratospheric injection of fuel results in cases where the fuel is not contained after re-entry, an extensive discussion of the fall-out mechanism is presented. (auth)

Dix, G.P.

1960-06-30T23:59:59.000Z

14

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

SciTech Connect

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.

Not Available

1988-05-01T23:59:59.000Z

15

Light-Weight Radioisotope Heater Unit Final Safety Analysis Report (LWRHU FSAR): Volume 3, Nuclear Risk Analysis Document  

SciTech Connect

The Light-Weight Radioisotope Heater Unit (LWRHU) Final Safety Analysis Report (FSAR), Volume 2, Accident Model Document (AMD) describes potential accident scenarios during the Galileo mission and evaluates the response of the LWRHUs to the associated accident environments. Any resulting source terms, consisting of PuO2 (with Pu-238 the dominant radionuclide), are then described in terms of curies released, particle size distribution, release location, and probabilities. This volume (LWRHU-FSAR, Volume 3, Nuclear Risk Analysis Document (NRAD)) contains the radiological analyses which estimate the consequences of the accident scenarios described in the AMD. It also contains the quantification of mission risks resulting from the LWRHUs based on consideration of all accident scenarios and their probabilities. Estimates of source terms and their characteristics derived in the AMD are used as inputs to the analyses in the NRAD. The Failure Abort Sequence Trees (FASTs) presented in the AMD define events for which source terms occur and quantify them. Based on this information, three types of source term cases (most probable, maximum, and expectation) for each mission phase were developed for use in evaluating the radiological consequences and mission risks. 4 refs., 5 figs., 8 tabs.

Not Available

1988-11-30T23:59:59.000Z

16

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

Science Conference Proceedings (OSTI)

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.

NONE

1994-10-01T23:59:59.000Z

17

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

Science Conference Proceedings (OSTI)

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.

TOMASZEWSKI, T.A.

2000-04-25T23:59:59.000Z

18

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

DOE Green Energy (OSTI)

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.

Not Available

1989-01-25T23:59:59.000Z

19

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

DOE Green Energy (OSTI)

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. 4 refs., 11 figs., 31 tabs.

Not Available

1989-01-13T23:59:59.000Z

20

Documented Safety Analysis  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Documented Safety Analysis Documented Safety Analysis FUNCTIONAL AREA GOAL: A document that provides an adequate description of the hazards of a facility during its design, construction, operation, and eventual cleanup and the basis to prescribe operating and engineering controls through Technical Safety Requirements (TSR) or Administrative Controls (AC). REQUIREMENTS:  10 CFR 830.204, Nuclear Safety Rule  DOE-STD-1027-92, Hazard Categorization, 1992.  DOE-STD-1104-96, Change Notice 1, Review and Approval of Nuclear Facility Safety Basis Documents (documented Safety Analyses and Technical Safety Requirements), dated May 2002.  DOE-STD-3009-2002, Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, Change Notice No. 2, April 2002.

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

SciTech Connect

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.

Not Available

1993-04-01T23:59:59.000Z

22

K Basin safety analysis  

DOE Green Energy (OSTI)

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.

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

1994-12-16T23:59:59.000Z

23

MACCS2 Final Gap Analysis  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MACCS2-Gap Analysis MACCS2-Gap Analysis Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.3: Software Quality Assurance Improvement Plan: MACCS2 Gap Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 MACCS2 Gap Analysis May 2004 Final Report INTENTIONALLY BLANK ii MACCS2 Gap Analysis May 2004 Final Report FOREWORD This report documents the outcome of an evaluation of the Software Quality Assurance (SQA) attributes of the radiological dispersion computer code, MACCS2, relative to established software requirements. This evaluation, a "gap analysis", is performed to meet commitment 4.2.1.3 of the

24

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

SciTech Connect

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

NONE

1995-02-01T23:59:59.000Z

25

Safety Analysis, Hazard and Risk Evaluations [Nuclear Waste Management  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety Analysis, Hazard Safety Analysis, Hazard and Risk Evaluations Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Safety Analysis, Hazard and Risk Evaluations Bookmark and Share NE Division personnel had a key role in the creation of the FCF Final Safety Analysis Report (FSAR), FCF Technical Safety Requirements (TSR)

26

FINAL ENVIRONMENTAL ASSESSMENT FOR ENVIRONMENTAL SAFETY AND  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

- - FOR ENVIRONMENTAL SAFETY AND HEALTH ANALYTICAL LABORATORY PROJECT NO. 94-AA-01 PANTEX PLANT AMAFsLLo, TEXAS June 1995 U.S. Department of E n e r g y Albuquerque Operations office Amarillo Area Office Pantex Plant P.O. Box 30030 Amarillo, Texas 79120 DISTRl6UTlON OF THIS DOCUMENT IS CMLtMmD FINAL ENVIRONMENTAL ASSESSMENT FOR June 1995 1 I U.S. Department of Energy Albuquerque Operations Office Amarillo Area m i c e Pantex Plant P.O. Box 30030 Amarillo, Texas 79120 DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. . --.-- - . . _ . I . . . . , . . . . . . . . . . - I I I TABLE OF CONTENTS m EXECUTIVESUMMARY .............................................. 1 1.0 PURPOSE AND NEED FOR AGENCY ACTION

27

Aviation safety analysis  

E-Print Network (OSTI)

Introduction: Just as the aviation system is complex and interrelated, so is aviation safety. Aviation safety involves design of aircraft and airports, training of ground personnel and flight crew members' maintenance of ...

Ausrotas, Raymond A.

1984-01-01T23:59:59.000Z

28

Ferrocyanide safety project ferrocyanide aging studies. Final report  

Science Conference Proceedings (OSTI)

This final report gives the results of the work conducted by Pacific Northwest National Laboratory (PNNL) from FY 1992 to FY 1996 on the Ferrocyanide Aging Studies, part of the Ferrocyanide Safety Project. The Ferrocyanide Safety Project was initiated as a result of concern raised about the safe storage of ferrocyanide waste intermixed with oxidants, such as nitrate and nitrite salts, in Hanford Site single-shell tanks (SSTs). In the laboratory, such mixtures can be made to undergo uncontrolled or explosive reactions by heating dry reagents to over 200{degrees}C. In 1987, an Environmental Impact Statement (EIS), published by the U.S. Department of Energy (DOE), Final Environmental Impact Statement, Disposal of Hanford Defense High-Level Transuranic and Tank Waste, Hanford Site, Richland, Washington, included an environmental impact analysis of potential explosions involving ferrocyanide-nitrate mixtures. The EIS postulated that an explosion could occur during mechanical retrieval of saltcake or sludge from a ferrocyanide waste tank, and concluded that this worst-case accident could create enough energy to release radioactive material to the atmosphere through ventilation openings, exposing persons offsite to a short-term radiation dose of approximately 200 mrem. Later, in a separate study (1990), the General Accounting Office postulated a worst-case accident of one to two orders of magnitude greater than that postulated in the DOE EIS. The uncertainties regarding the safety envelope of the Hanford Site ferrocyanide waste tanks led to the declaration of the Ferrocyanide Unreviewed Safety Question (USQ) in October 1990.

Lilga, M.A.; Hallen, R.T.; Alderson, E.V. [and others

1996-06-01T23:59:59.000Z

29

Light-Weight Radioisotope Heater Unit final safety analysis report (LWRHU-FSAR): Volume 2: Accident Model Document (AMD)  

Science Conference Proceedings (OSTI)

The purpose of this volume of the LWRHU SAR, the Accident Model Document (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; Provide estimates of occurrence probabilities associated with these various accidents; Evaluate the response of the LWRHU (or its components) to the resultant accident environments; and Associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere.

Johnson, E.W.

1988-10-01T23:59:59.000Z

30

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

SciTech Connect

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

T. J. Rodovsky

2006-03-01T23:59:59.000Z

31

Waste Isolation Pilot Plant Safety Analysis Report  

Science Conference Proceedings (OSTI)

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.

NONE

1995-11-01T23:59:59.000Z

32

Software Quality Assurance Improvment Plan: ALOHA Gap Analysis, Final Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Final-ALOHA Final-ALOHA Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.3: Software Quality Assurance Improvement Plan: ALOHA Gap Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 ALOHA Gap Analysis May 2004 Final Report INTENTIONALLY BLANK ii ALOHA Gap Analysis May 2004 Final Report FOREWORD This report documents the outcome of an evaluation of the Software Quality Assurance (SQA) attributes of the chemical source term and atmospheric dispersion computer code, ALOHA 5.2.3, relative to established

33

Molten salt safety study. Final report  

DOE Green Energy (OSTI)

The considerations concerning safety in using molten salt (40% potassium nitrate, 60% sodium nitrate) in a solar central receiver plant are addressed. The considerations are of a general nature and do not cover any details of equipment or plant operation. The study includes salt chemical reaction, experiments with molten salt, dry storage and handling constraints, and includes data from the National Fire Protection Association. The contents of this report were evaluated by two utility companies and they concluded that no major safety problems exist in using a molten salt solar system.

Not Available

1980-01-01T23:59:59.000Z

34

N reactor safety upgrades final status report  

Science Conference Proceedings (OSTI)

This document describes the requirements, objectives, work completed, and issues resolved for all safety upgrades recommended by the various expert panels (e.g., Roddis), National Academy of Sciences (NAS), and Westinghouse Independent Safety Appraisal (WISA) team. This report reflects the status of the recommendations at the time of publication, while the reactor is being placed in cold standby. Each recommendation required a resolution, a closure process if closed, and a punchlist reference if open and required for restart. These recommendations are maintained on databases that were updated periodically as work was completed or punchlisted. The databases and files are to be retained with the N Reactor Standby files.

Foreman, S.K.; Rainey, T.E.; Erpenbeck, E.G.

1990-02-01T23:59:59.000Z

35

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

NLE Websites -- All DOE Office Websites (Extended Search)

(for facility modifications or at LSO's direction), and potential inadequacy of the safety analysis (PISA) submittals. The process for DSA and TSR reviews is detailed and...

36

Hotspot Gap Analysis Final 20070323  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

HS-0003 HS-0003 Software Evaluation of Hotspot and DOE Safety Software Toolbox Recommendation U.S. Department of Energy Office of Health, Safety and Security 1000 Independence Avenue, S.W. Washington, DC 20585-2040 March, 2007 ii Foreword This report documents the outcome of an evaluation of the Safety Software Quality Assurance (SSQA) attributes of Hotspot, a health physics application, relative to the safety software requirements identified in DOE O 414.1C, Quality Assurance. This evaluation, a "gap analysis", is performed according to the implementation guide DOE G 414.1-4, and is a requisite for deciding whether Hotspot should be designated as a toolbox code for DOE's safety software Central Registry. Comments regarding this document should be addressed to:

37

Safety analysis of the UTSI-CFFF superconducting magnet  

DOE Green Energy (OSTI)

In designing a large superconducting magnet such as the UTSI-CFFF dipole, great attention must be devoted to the safety of the magnet and personnel. The conductor for the UTSI-CFFF magnet incorporates much copper stabilizer, which both insures its cryostability, and contributes to the magnet safety. The quench analysis and the cryostat fault condition analysis are presented. Two analyses of exposed turns follow; the first shows that gas cooling protects uncovered turns; the second, that the cryostat pressure relief system protects them. Finally the failure mode and safety analysis is presented.

Turner, L.R.; Wang, S.T.; Smith, R.P.; VanderArend, P.C.; Hsu, Y.H.

1979-01-01T23:59:59.000Z

38

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

SciTech Connect

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

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

1994-02-01T23:59:59.000Z

39

Solid waste burial grounds interim safety analysis  

SciTech Connect

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.

Saito, G.H.

1994-10-01T23:59:59.000Z

40

Software Quality Assurance Improvment Plan: CFAST Gap Analysis, Final Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EH-4.2.1.3-CFAST-Gap Analysis EH-4.2.1.3-CFAST-Gap Analysis Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.3: Software Quality Assurance Improvement Plan: CFAST Gap Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 CFAST Gap Analysis May 2004 Final Report ii INTENTIONALLY BLANK CFAST Gap Analysis May 2004 Final Report iii FOREWORD This report documents the outcome of an evaluation of the Software Quality Assurance (SQA) attributes of the CFAST computer code for accident analysis applications, relative to established requirements. This evaluation, a "gap analysis," is performed to meet commitment 4.2.1.3 of the Department of Energy's

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Hot Cell Facility (HCF) Safety Analysis Report  

Science Conference Proceedings (OSTI)

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.

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

2000-11-01T23:59:59.000Z

42

Nuclear Criticality Safety Application Guide: Safety Analysis Report Update Program  

SciTech Connect

Martin Marietta Energy Systems, Inc. (MMES) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Safety analyses are performed to identify hazards and potential accidents; to analyze the adequacy of measures taken to eliminate, control, or mitigate hazards; and to evaluate potential accidents and determine associated risks. Safety Analysis Reports (SARs) are prepared to document the safety analysis to ensure facilities can be operated safely and in accordance with regulations. Many of the facilities requiring a SAR process fissionable material creating the potential for a nuclear criticality accident. MMES has long had a nuclear criticality safety program that provides the technical support to fissionable material operations to ensure the safe processing and storage of fissionable materials. The guiding philosophy of the program has always been the application of the double-contingency principle, which states: {open_quotes}process designs shall incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.{close_quotes} At Energy Systems analyses have generally been maintained to document that no single normal or abnormal operating conditions that could reasonably be expected to occur can cause a nuclear criticality accident. This application guide provides a summary description of the MMES Nuclear Criticality Safety Program and the MMES Criticality Accident Alarm System requirements for inclusion in facility SARs. The guide also suggests a way to incorporate the analyses conducted pursuant to the double-contingency principle into the SAR. The prime objective is to minimize duplicative effort between the NCSA process and the SAR process and yet adequately describe the methodology utilized to prevent a nuclear criticality accident.

1994-02-01T23:59:59.000Z

43

Microsoft Word - IMBA Gap Analysis Final 20060831.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE/EH-0711 DOE/EH-0711 Gap Analysis for IMBA and DOE Safety Software Central Registry Recommendation Final U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Avenue, S.W. Washington, D.C. 20585-2040 August 2006 DOE/EH-0711 i INTENTIONALLY BLANK DOE/EH-0711 ii FOREWORD This report documents the outcome of an evaluation of the safety software quality assurance attributes of the Integrated Modules for Bioassay Analysis (IMBA) Expert (tm) USDOE-Edition and Professional Plus computer products relative to the safety software requirements identified in DOE O 414.1C, Quality Assurance. This evaluation, a gap analysis, is performed according to DOE G 414.1-4 and is a requisite

44

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

are anticipated, so this authority is not currently needed. The LSO Integrated Safety Management System Description and Environment, Safety and Health Functions,...

45

Washington TRU Solutions - Job Safety/Hazard Analysis Booklet  

NLE Websites -- All DOE Office Websites (Extended Search)

participation is important to efficient, safe, and increased productivity. Through the process of Job Safety Analysis, these benefits are fully realized. Job Safety Analysis can...

46

Using Addenda in Documented Safety Analysis Reports  

Science Conference Proceedings (OSTI)

This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update.

Douglas S. Swanson; Michael A. Thieme

2003-06-01T23:59:59.000Z

47

Using Addenda in Documented Safety Analysis Reports  

Science Conference Proceedings (OSTI)

This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update.

Swanson, D.S.; Thieme, M.A.

2003-06-16T23:59:59.000Z

48

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

Energy.gov (U.S. Department of Energy (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.

49

Seismic Safety Margins Research Program. Phase I, final report - overview  

SciTech Connect

The Seismic Safety Margins Research Program (SSMRP) is a multiyear, multiphase program whose overall objective is to develop improved methods for seismic safety assessments of nuclear power plants, using a probabilistic computational procedure. The program is being carried out at the Lawrence Livermore National Laboratory and is sponsored by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Phase I of the SSMRP was successfully completed in January 1981: A probabilistic computational procedure for the seismic risk assessment of nuclear power plants has been developed and demonstrated. The methodology is implemented by three computer programs: HAZARD, which assesses the seismic hazard at a given site, SMACS, which computes in-structure and subsystem seismic responses, and SEISIM, which calculates system failure probabilities and radioactive release probabilities, given (1) the response results of SMACS, (2) a set of event trees, (3) a family of fault trees, (4) a set of structural and component fragility descriptions, and (5) a curve describing the local seismic hazard. The practicality of this methodology was demonstrated by computing preliminary release probabilities for Unit 1 of the Zion Nuclear Power Plant north of Chicago, Illinois. Studies have begun aimed at quantifying the sources of uncertainty in these computations. Numerous side studies were undertaken to examine modeling alternatives, sources of error, and available analysis techniques. Extensive sets of data were amassed and evaluated as part of projects to establish seismic input parameters and to produce the fragility curves. 66 refs., 29 figs., 10 tabs.

Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Chuang, T.Y.; Cummings, G.E.; Johnson, J.J.; Mensing, R.W.; Wells, J.E.

1981-03-06T23:59:59.000Z

50

Microsoft Word - Nuclear Safety Pamphlet Final September 1 2010...  

NLE Websites -- All DOE Office Websites (Extended Search)

A Basic Overview of NUCLEAR SAFETY AT THE DEPARTMENT OF ENERGY Outreach & Awareness Series Office of Health, Safety and Security (HSS) U.S. Department of Energy September 2010...

51

Safety of ncker's strictness analysis  

Science Conference Proceedings (OSTI)

This paper proves correctness of Ncker's method of strictness analysis, implemented in the Clean compiler, which is an effective way for strictness analysis in lazy functional languages based on their operational semantics. We improve upon the ...

Manfred Schmidt-schauss; David Sabel; Marko Schtz

2008-07-01T23:59:59.000Z

52

K West integrated water treatment system subproject safety analysis document  

Science Conference Proceedings (OSTI)

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.

SEMMENS, L.S.

1999-02-24T23:59:59.000Z

53

Fire safety of LPG in marine transportation. Final report  

SciTech Connect

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

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

1980-06-01T23:59:59.000Z

54

Final Technical Report on Radioxenon Event Analysis  

SciTech Connect

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

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

2013-03-15T23:59:59.000Z

55

SYNTHESIS OF SAFETY ANALYSIS AND FIRE HAZARD ANALYSIS METHODOLOGIES  

Science Conference Proceedings (OSTI)

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.

Coutts, D

2007-04-17T23:59:59.000Z

56

Technical Standards, Safety Analysis Toolbox Codes - November 2003 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Analysis Toolbox Codes - November 2003 Safety Analysis Toolbox Codes - November 2003 Technical Standards, Safety Analysis Toolbox Codes - November 2003 November 2003 Software Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes Safety analysis software for the DOE "toolbox" was designated by DOE/EH in March 2003 (DOE/EH, 2003). The supporting basis for this designation was provided by a DOE-chartered Safety Analysis Software Group in the technical report, Selection of Computer Codes for DOE Safety Analysis Applications, (August, 2002). Technical Standards, Safety Analysis Toolbox Codes More Documents & Publications DOE G 414.1-4, Safety Software Guide for Use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance Technical Standards, MELCOR - Gap Analysis - May 3, 2004

57

Operation Castle. Radiological Safety. Volume 2. Final report  

SciTech Connect

This report is designed to cover the overall Operation Castle radiological safety matters from the viewpoint of those issues of direct concern to Headquarters, Joint Task Force Seven. It was written for the express purpose of assisting in the development of future radiological safety plans by presenting detailed discussion of the problems and solutions arising during Operation Castle.

Not Available

1985-09-01T23:59:59.000Z

58

Operation Castle. Radiological Safety. Volume 1. Final report  

SciTech Connect

This report is designed to cover the overall Operation Castle radiological safety matters from the viewpoint of those issues of direct concern to Headquarters, Joint Task Force Seven. It was written for the express purpose of assisting in the development of future radiological safety plans by presenting detailed discussion of the problems and solutions arising during Operation Castle. Included is a discussion of fallout forecasting techniques.

Not Available

1985-09-01T23:59:59.000Z

59

Nuclear Safety Management, Final Rule; Delay of Effective Date (66 FR  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Safety Management, Final Rule; Delay of Effective Date (66 Nuclear Safety Management, Final Rule; Delay of Effective Date (66 FR 8746), Fed Reg, 2/2/01 Nuclear Safety Management, Final Rule; Delay of Effective Date (66 FR 8746), Fed Reg, 2/2/01 Nuclear Safety Management, Final Rule; Delay of Effective Date (66 FR 8746), Fed Reg, 2/2/01 In accordance with the memorandum of January 20, 2001, from the Assistant to the President and Chief of Staff, entitled ''Regulatory Review Plan,'' published in the Federal Register on January 24, 2001 (66 FR 7702), this action temporarily delays for 60 days the effective date of the rule entitled ''Alternate Fuel Transportation Program; Biodiesel Fuel Use Credit'' published in the Federal Register on January 11, 2001 (66 FR 2207). DATES: The effective date of the rule amending 10 CFR part 490

60

Nuclear Safety Management, Final Rule amending 10 CFR Part 830 (66 FR  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management, Final Rule amending 10 CFR Part 830 (66 Management, Final Rule amending 10 CFR Part 830 (66 FR 1810), Federal Register (Fed Reg), 1/10/2001 Nuclear Safety Management, Final Rule amending 10 CFR Part 830 (66 FR 1810), Federal Register (Fed Reg), 1/10/2001 SUMMARY: The Department of Energy (DOE) adopts, with minor changes, the interim final rule published on October 10, 2000, to amend the DOE Nuclear Safety Management regulations. EFFECTIVE DATE: This final rule is effective on February 9, 2001. FOR FURTHER INFORMATION CONTACT: Richard Black, Director, Office of Nuclear and Facility Safety Policy, 270CC, Department of Energy, 19901 Germantown Road, Germantown, MD 20874; telephone: 301-903-3465; email: Richard.Black@eh.doe.gov SUPPLEMENTARY INFORMATION: I. Introduction and Summary On October 10, 2000, the Department of Energy (DOE) published an

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Safety Bulletin 2011-01, Events Beyond Design Safety Basis Analysis |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Bulletin 2011-01, Events Beyond Design Safety Basis Analysis Bulletin 2011-01, Events Beyond Design Safety Basis Analysis Safety Bulletin 2011-01, Events Beyond Design Safety Basis Analysis PURPOSE This Safety Alert provides information on a safety concern related to the identification and mitigation of events that may fall outside those analyzed in the documented safety analysis. BACKGROUND On March 11 , 2011 , the Fukushima Daiichi nuclear power station in Japan was damaged by a magnitude 9.0 earthquake and the subsequent tsunami. While there is still a lot to be learned from the accident · about the adequacy of design specifications and the equipment failure modes, reports from the Nuclear Regulatory Commission (NRC) have identified some key aspects of the operational emergency at the Fukushima Daiichi nuclear power station.

62

NVLAP gap analysis. Final report  

SciTech Connect

The capabilities of AlliedSignal Metrology were compared to the requirements contained in the National Voluntary Laboratory Accreditation Program (NVLAP) Calibration Laboratories Handbook and NVLAP Calibration Laboratories Technical Guide. The initial analyses demonstrated a need for improved measurement uncertainty analyses, additional control artifacts, and improved statistical process control. The analysis also revealed the need for a formalized customer complaint system and a calibration quality manual.

Shroyer, K.A.

1997-04-01T23:59:59.000Z

63

Package Safety Analysis Assessment for Sludge Transportation System  

SciTech Connect

This package safety analysis assessment demonstrates that the Sludge Transportation System meets the acceptance criteria for an equivalent package as specified in DOE/RL-2001-36, Hanford Sitewide Transportation Safety Document for onsite shipment.

ROMANO, T.

2003-03-19T23:59:59.000Z

64

Mechanistic facility safety and source term analysis  

SciTech Connect

A PC-based computer program was created for facility safety and source term analysis at Hanford The program has been successfully applied to mechanistic prediction of source terms from chemical reactions in underground storage tanks, hydrogen combustion in double contained receiver tanks, and proccss evaluation including the potential for runaway reactions in spent nuclear fuel processing. Model features include user-defined facility room, flow path geometry, and heat conductors, user-defined non-ideal vapor and aerosol species, pressure- and density-driven gas flows, aerosol transport and deposition, and structure to accommodate facility-specific source terms. Example applications are presented here.

PLYS, M.G.

1999-06-09T23:59:59.000Z

65

LPG land transportation and storage safety. Final report  

SciTech Connect

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

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

1981-09-01T23:59:59.000Z

66

LPG land transportation and storage safety. Final report  

SciTech Connect

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

1981-09-01T23:59:59.000Z

67

Review of Documented Safety Analysis Development for the Hanford...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. Department of Energy Subject: Review of Documented Safety Analysis Development for the Hanford Site Waste Treatment and Immob ilization Plant (LBL Facilities) - C riteria and...

68

Aviation Safety and Air Traffic Management Analysis - Center...  

NLE Websites -- All DOE Office Websites (Extended Search)

support tools. Visualization and analysis of diverse data sources including flight track, weather, airport, aircraft, ATM elements and geographic data supports aviation safety...

69

CRAD, Documented Safety Analysis Development - April 23, 2013 | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Documented Safety Analysis Development - April 23, 2013 Documented Safety Analysis Development - April 23, 2013 CRAD, Documented Safety Analysis Development - April 23, 2013 April 23, 2013 Review of Documented Safety Analysis Development for the Hanford Site Waste Treatment and Immobilization Plant (LBL Facilities) (HSS CRAD 45-58, Rev. 0) The review will consider selected aspects of the development of the Documented Safety Analysis (DSA) for the Waste Treatment and Immobilization Plant (WTP); Low Activity Waste (LAW) facility, Balance of Facilities and Analytical Laboratory (LAB) (collectively identified as LBL) to assess the extent to which nuclear safety is integrated into the design of the LBL facilities in accordance with DOE directives; in particular, DOE Order 420. l B and DOE-STD-3009-94. The review will focus on a few selected

70

Volume II - Accident and Operational Safety Analysis Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

208-2012 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 (EFCOG), Industrial Hygiene and Safety Sub-group of the Environmental Health and Safety Working Group. The preparers would like to gratefully acknowledge the authors whose works are referenced in this document, and the individuals who provided valuable technical insights and/or specific

71

Safety Evaluation Report of the Waste Isolation Pilot Plant Contact Handled (CH) Waste Documented Safety Analysis  

Science Conference Proceedings (OSTI)

This Safety Evaluation Report (SER) documents the Department of Energys (DOE's) review of Revision 9 of the Waste Isolation Pilot Plant Contact Handled (CH) Waste Documented Safety Analysis, DOE/WIPP-95-2065 (WIPP CH DSA), and provides the DOE Approval Authority with the basis for approving the document. It concludes that the safety basis documented in the WIPP CH DSA is comprehensive, correct, and commensurate with hazards associated with CH waste disposal operations. The WIPP CH DSA and associated technical safety requirements (TSRs) were developed in accordance with 10 CFR 830, Nuclear Safety Management, and DOE-STD-3009-94, Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports.

Washington TRU Solutions LLC

2005-09-01T23:59:59.000Z

72

Final Rule for Nuclear Safety Management (10 CFR Part 830)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

717 717 Federal Register / Vol. 66, No. 74 / Tuesday, April 17, 2001 / Rules and Regulations engineering and cost analyses because the results showed that the two blowing agent alternatives can be used to achieve similar performance for similar costs to HFC-245fa. DOE estimates are reasonable and address the concern of the Department of Justice to provide more than one choice of insulation blowing agent with comparable performance and at approximately the same cost. Based on the analysis of the three different types of blowing agents, HFC- 245fa-, pentane/cyclopentane- and HFC-134a, DOE concluded that water heater manufacturers will have several choices to reach the standard, including blends of these blowing agents, and therefore, will not have to rely on a sole source supplier.

73

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

Science Conference Proceedings (OSTI)

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.

WILLIAMS, J.C.

2003-11-15T23:59:59.000Z

74

Accident analysis and safety review of DOE Category B reactors  

SciTech Connect

DOE is employing the principle of comparability with the NRC requirements to guide its safety program. Since the safety record of research reactors licensed by the NRC has been established and accepted, the comparison of DOE Orders applicable to DOE research reactors with the NRC regulations applicable to research reactors would identify strengths and weaknesses of the DOE Orders. The comparison was made in 14 general topics of safety which are labeled Areas of Safety Concerns. This paper focuses on the Area of accident analysis and safety review and presents recommendations in these areas. 12 refs.

Kimura, C.Y.

1990-08-07T23:59:59.000Z

75

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11 Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11 During the 580th meeting of the Advisory Committee on Reactor Safeguards (ACRS), February 10-12, 2011, we reviewed the staff's white paper, "A Comparison of Integrated Safety Analysis and Probabilistic Risk Assessment." Our Radiation Protection and Nuclear Materials Subcommittee also reviewed this matter during a meeting on January 11, 2011. During these meetings we met with representatives of the NRC staff and the Nuclear Energy Institute. We also had the benefit of the documents referenced. Comparison of Intergrated Safety Analysis (ISA) and Probabilistic Risk

76

Regulatory analysis technical evaluation handbook. Final report  

Science Conference Proceedings (OSTI)

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

NONE

1997-01-01T23:59:59.000Z

77

Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689Nuclear Energy Institute (NEI) Letter, 9/10/10 Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689Nuclear Energy Institute (NEI) Letter, 9/10/10 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 intended as an information source for the NRC and should serve as a foundation for discussion with industry representatives on the issue. This paper concludes that an ISA is a risk-informed, performance-based way of achieving and maintaining safety at fuel recycling facilities. As

78

Safety analysis report for the Waste Storage Facility. Revision 2  

SciTech Connect

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.

Bengston, S.J.

1994-05-01T23:59:59.000Z

79

SNF fuel retrieval sub project safety analysis document  

SciTech Connect

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.

BERGMANN, D.W.

1999-02-24T23:59:59.000Z

80

Planning Document for an NBSR Conversion Safety Analysis Report  

SciTech Connect

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.

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

2013-09-25T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Evolutionary safety analysis: motivations from the air traffic management domain  

E-Print Network (OSTI)

Abstract. In order realistically and cost-effectively to realize the ATM (Air Traffic Management) 2000+ Strategy, systems from different suppliers will be interconnected to form a complete functional and operational environment, covering ground segments and aerospace. Industry will be involved as early as possible in the lifecycle of ATM projects. EURO-CONTROL manages the processes that involve the definition and validation of new ATM solutions using Industry capabilities (e.g., SMEs). In practice, safety analyses adapt and reuse system design models (produced by third parties). Technical, organisational and cost-related reasons often determine this choice, although design models are unfit for safety analysis. Design models provide limited support to safety analysis, because they are tailored for system designers. The definition of an adequate model and of an underlying methodology for its construction will be highly beneficial for whom is performing safety analyses. Limited budgets and resources, often, constrain or inhibit the model definition phase as an integral part of safety analysis. This paper is concerned with problems in modeling ATM systems for safety analysis. The main objective is to highlight a model specifically targeted to support evolutionary safety analysis. 1

Massimo Felici

2005-01-01T23:59:59.000Z

82

Final Report - Hydrogen Delivery Infrastructure Options Analysis  

NLE Websites -- All DOE Office Websites (Extended Search)

The Power of Experience The Power of Experience Final Report Hydrogen Delivery Infrastructure Options Analysis DOE Award Number: DE-FG36-05GO15032 Project director/principal investigator: Tan-Ping Chen Consortium/teaming Partners: Air Liquide, Chevron Technology Venture, Gas Technology Institute, NREL, Tiax, ANL Hydrogen Delivery Infrastructure Options Analysis ii TABLE OF CONTENTS SECTION 1 EXECUTIVE SUMMARY ........................................................................... 1-1 1.1 HOW THE RESEARCH ADDS TO THE UNDERSTANDING OF THE AREA INVESTIGATED. 1-1 1.2 TECHNICAL EFFECTIVENESS AND ECONOMIC FEASIBILITY OF THE METHODS OR TECHNIQUES INVESTIGATED OR DEMONSTRATED .................................................... 1-1 1.3 HOW THE PROJECT IS OF BENEFIT TO THE PUBLIC..................................................... 1-1

83

Computational methods for criticality safety analysis within the scale system  

SciTech Connect

The criticality safety analysis capabilities within the SCALE system are centered around the Monte Carlo codes KENO IV and KENO V.a, which are both included in SCALE as functional modules. The XSDRNPM-S module is also an important tool within SCALE for obtaining multiplication factors for one-dimensional system models. This paper reviews the features and modeling capabilities of these codes along with their implementation within the Criticality Safety Analysis Sequences (CSAS) of SCALE. The CSAS modules provide automated cross-section processing and user-friendly input that allow criticality safety analyses to be done in an efficient and accurate manner. 14 refs., 2 figs., 3 tabs.

Parks, C.V.; Petrie, L.M.; Landers, N.F.; Bucholz, J.A.

1986-01-01T23:59:59.000Z

84

Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Energy Institute (NEI) Attachment, Integrated Safety Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis This paper addresses why the use of an Integrated Safety Analysis ("ISA") is appropriate for fuel recycling facilities1 which would be licensed under new regulations currently being considered by NRC. The use of the ISA for fuel facilities under Part 70 is described and compared to the use of a Probabilistic Risk Assessment ("PRA") for reactor facilities. A basis is provided for concluding that future recycling facilities - which will possess characteristics similar to today's fuel cycle facilities and distinct from reactors - can best be assessed using established qualitative or semi-quantitative ISA techniques to achieve and

85

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

Science Conference Proceedings (OSTI)

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

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

1994-02-01T23:59:59.000Z

86

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

Science Conference Proceedings (OSTI)

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

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

2011-12-15T23:59:59.000Z

87

Process hazards analysis (PrHA) program, bridging accident analyses and operational safety  

SciTech Connect

Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker safety are incorporated so the worker can readily identify the safety parameters of the their work. System safety tools such as Preliminary Hazard Analysis, What-If Analysis, Hazard and Operability Analysis as well as other techniques as necessary provide the groundwork for both determining bounding conditions for facility safety, operational safety, and day-to-clay worker safety.

Richardson, J. A. (Jeanne A.); McKernan, S. A. (Stuart A.); Vigil, M. J. (Michael J.)

2003-01-01T23:59:59.000Z

88

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

Science Conference Proceedings (OSTI)

This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety Implications of Advanced Technology Power Conversion and Design Innovations and Simplifications: Investigations of supercritical CO{sub 2} gas turbine Brayton cycles coupled to the sodium-cooled reactors and innovative concepts for sodium-to-CO{sub 2} heat exchangers were performed to discover new designs for high efficiency electricity production. The objective of the analyses was to characterize the design and safety performance of equipment needed to implement the new power cycle. The project included considerations of heat transfer and power conversion systems arrangements and evaluations of systems performance. Task 4--Post Accident Heat Removal and In-Vessel Retention: Test plans were developed to evaluate (1) freezing and plugging of molten metallic fuel in subassembly geometry, (2) retention of metallic fuel core melt debris within reactor vessel structures, and (3) consequences of intermixing of high pressure CO{sub 2} and sodium. The objective of the test plan development was to provide planning for measurements of data needed to characterize the consequences of very low probability accident sequences unique to metallic fuel and CO{sub 2} Brayton power cycles. The project produced three test plans ready for execution.

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

2007-08-13T23:59:59.000Z

89

Hazard Analysis Database report  

Science Conference Proceedings (OSTI)

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

Niemi, B.J.

1997-08-12T23:59:59.000Z

90

Safety analysis report for packaging (onsite) steel drum  

SciTech Connect

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.

McCormick, W.A.

1998-09-29T23:59:59.000Z

91

Safety Analysis Of Automated Highway Systems  

E-Print Network (OSTI)

hazard analysis of process-control software requirements. In contrast, the analyst does not have a tool

Leveson, Nancy G.

1997-01-01T23:59:59.000Z

92

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

SciTech Connect

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

Fisk, Patricia; Rutherford, Lavon

2003-06-01T23:59:59.000Z

93

Final safety analysis report for the irradiated fuels storage facility  

SciTech Connect

A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1$sup 1$/$sub 2$ cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100$sup 0$F is reached. (LK)

Bingham, G.E.; Evans, T.K.

1976-01-01T23:59:59.000Z

94

Hanford Sludge Treatment Project 105-KW Final Safety Analysis...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report (HNF-SD-WM-SAR-062, Revision 14C) for the Sludge Treatment Project at the Hanford Site. 2.0 BACKGROUND The Sludge Treatment Project manages the removal of radioactive...

95

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

SciTech Connect

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

Miller, G.L.

1997-01-20T23:59:59.000Z

96

Applicability of trends in nuclear safety analysis to space nuclear power systems  

SciTech Connect

A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication.

Bari, R.A.

1992-10-01T23:59:59.000Z

97

QuantUM: Quantitative Safety Analysis of UML Models  

E-Print Network (OSTI)

When developing a safety-critical system it is essential to obtain an assessment of different design alternatives. In particular, an early safety assessment of the architectural design of a system is desirable. In spite of the plethora of available formal quantitative analysis methods it is still difficult for software and system architects to integrate these techniques into their every day work. This is mainly due to the lack of methods that can be directly applied to architecture level models, for instance given as UML diagrams. Also, it is necessary that the description methods used do not require a profound knowledge of formal methods. Our approach bridges this gap and improves the integration of quantitative safety analysis methods into the development process. All inputs of the analysis are specified at the level of a UML model. This model is then automatically translated into the analysis model, and the results of the analysis are consequently represented on the level of the UML model. Thus the analysi...

Leitner-Fischer, Florian; 10.4204/EPTCS.57.2

2011-01-01T23:59:59.000Z

98

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

Science Conference Proceedings (OSTI)

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.

Kerr, C E

2009-10-07T23:59:59.000Z

99

Safety and Techno-Economic Analysis of Solvent Selection for Supercritical Fischer-Tropsch Synthesis Reactors  

E-Print Network (OSTI)

Fisher-Tropsch Synthesis is a primary pathway for gas-to-liquid technology. In order to overcome commercial problems associated with reaction and transport phenomena, the use of supercritical solvents has been proposed to increase chemical conversion and improve temperature control. One of the major challenges in designing the supercritical FTS systems is the solvent selection. Numerous alternatives exist and should be screened based on relevant criteria. The main aim of the thesis was to develop a safety metric that can be incorporated in the selection of an optimal supercritical solvent or a mixture of solvents. The objective was to minimize the cost while satisfying safety constraints or to establish tradeoffs between cost and safety. Hydrocarbons from C3 to C9 were identified as feasible solvents for FTS purposes. The choice of these solvents is dependent on their mixture critical temperature and pressure requirements that need to be satisfied upon entry into the FTS reactor. A safety metric system was developed in order to compare the risk issues associated with using the aforementioned solvents. In addition, an economic analysis of using the different solvents was performed. Finally, a case study was solved to illustrate the use of the proposed metrics and the selection of solvents based on safety and techno-economic criteria.

Hamad, Natalie

2011-12-01T23:59:59.000Z

100

Style, content and format guide for writing safety analysis documents. Volume 1, Safety analysis reports for DOE nuclear facilities  

Science Conference Proceedings (OSTI)

The purpose of Volume 1 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Analysis Reports (SARs) for DOE nuclear facilities at Sandia National Laboratories. The scope of Volume 1 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SARs for DOE nuclear facilities.

Not Available

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Final safeguards analysis, High Temperature Lattice Test Reactor  

SciTech Connect

Information on the HTLTR Reactor is presented concerning: reactor site; reactor buildings; reactor kinetics and design characteristics; experimental and test facilitles; instrumentation and control; maintenance and modification; initial tests and operations; administration and procedural safeguards; accident analysis; seifterminated excursions; main heat exchanger leak; training program outline; and reliability analysis of safety systems. (7 references) (DCC)

Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

1966-01-01T23:59:59.000Z

102

EIS-0225-SA-02: Final Supplement Analysis | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2: Final Supplement Analysis 2: Final Supplement Analysis EIS-0225-SA-02: Final Supplement Analysis Continued Operation of the Pantex Plant and Associated Storage of Nuclear Weapons Components Hazardous Waste Treatment and Processing Facility This SA specifically addresses the issue of housing liquid processes in a separate building, the elimination of forklift airlocks and overhead hoists from the main HWTPF, the handling of classified material, and the construction of a ramp instead of a shipping dock. DOE/EIS-0225, Final Supplement Analysis for the Final Environmental Impact Statement for the Continued Operation of the Pantex Plant and Associated Storage of Nuclear Weapons Components Hazardous Waste Treatment and Processing Facility (January 2000) More Documents & Publications

103

PAT-1 safety analysis report addendum.  

SciTech Connect

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.

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.

2010-09-01T23:59:59.000Z

104

The Office of Health, Safety and Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Office Oversight of the Fire Protection Program at the Oak Ridge Reservation, August 2011 Hanford Sludge Treatment Project 105-KW Final Safety Analysis Report Review, August 2011...

105

Reactor Safety Testing and Analysis - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

106

Analysis Tools for Nuclear Criticality Safety - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

107

Methods and criteria for safety analysis (FIN L2535)  

SciTech Connect

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.

1992-12-01T23:59:59.000Z

108

Methods and criteria for safety analysis (FIN L2535)  

SciTech Connect

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.

1992-12-01T23:59:59.000Z

109

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

Science Conference Proceedings (OSTI)

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.

MCCOY, J.C.

1999-03-16T23:59:59.000Z

110

EA-1812: Final Supplement Analysis | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Final Supplement Analysis Final Supplement Analysis EA-1812: Final Supplement Analysis This Supplement Analysis (SA) has been prepared to address changes in the design and operating parameters of the NECO (formerly Haxtun) Wind Farm Project ("original proposed project") in Logan and Phillips Counties, Colorado. In January 2012, the DOE published the Final Environmental Assessment ("DOE/EA-1812") for the original proposed project and published a Finding of No Significant Impact (FONSI) on January 4, 2012. DOE/EA-1812 was conducted to analyze and disclose potential environmental and socioeconomic impacts that would result from the construction and operation of the original proposed project, which received federal funding through a Community Renewable Energy Deployment (CRED) Program grant to

111

Asbestos Exposure Limit AGENCY: Mine Safety and Health Administration, Labor. ACTION: Final rule. SUMMARY: The Mine Safety and Health  

E-Print Network (OSTI)

Administration (MSHA) is revising its existing health standards for asbestos exposure at metal and nonmetal mines, surface coal mines, and surface areas of underground coal mines. This final rule reduces the permissible exposure limits for airborne asbestos fibers and makes clarifying changes to the existing standards. Exposure to asbestos has been associated with lung cancer, mesothelioma, and other cancers, as well as asbestosis and other nonmalignant respiratory diseases. This final rule will help improve health protection for miners who work in an environment where asbestos is present and lower the risk that miners will suffer material impairment of health or functional capacity over their working lifetime. DATES: This final rule is effective April

Rwilkins On Prodpc Rules_; Patricia W. Silvey At

2008-01-01T23:59:59.000Z

112

Environmental Safety and Health Analytical Laboratory, Pantex Plant, Amarillo, Texas. Final Environmental Assessment  

SciTech Connect

The US Department of Energy (DOE) has prepared an Environmental Assessment (EA) of the construction and operation of an Environmental Safety and Health (ES&H) Analytical Laboratory and subsequent demolition of the existing Analytical Chemistry Laboratory building at Pantex Plant near Amarillo, Texas. In accordance with the Council on Environmental Quality requirements contained in 40 CFR 1500--1508.9, the Environmental Assessment examined the environmental impacts of the Proposed Action and discussed potential alternatives. Based on the analysis of impacts in the EA, conducting the proposed action, construction of an analytical laboratory and demolition of the existing facility, would not significantly effect the quality of the human environment within the meaning of the National Environmental Policy Act of 1969 (NEPA) and the Council on Environmental Quality regulations in 40 CFR 1508.18 and 1508.27.

NONE

1995-06-01T23:59:59.000Z

113

Safety analysis, 200 Area, Savannah River Plant: Separations area operations  

Science Conference Proceedings (OSTI)

The nev HB-Line, located on the fifth and sixth levels of Building 221-H, is designed to replace the aging existing HB-Line production facility. The nev HB-Line consists of three separate facilities: the Scrap Recovery Facility, the Neptunium Oxide Facility, and the Plutonium Oxide Facility. There are three separate safety analyses for the nev HB-Line, one for each of the three facilities. These are issued as supplements to the 200-Area Safety Analysis (DPSTSA-200-10). These supplements are numbered as Sup 2A, Scrap Recovery Facility, Sup 2B, Neptunium Oxide Facility, Sup 2C, Plutonium Oxide Facility. The subject of this safety analysis, the, Plutonium Oxide Facility, will convert nitrate solutions of {sup 238}Pu to plutonium oxide (PuO{sub 2}) powder. All these new facilities incorporate improvements in: (1) engineered barriers to contain contamination, (2) barriers to minimize personnel exposure to airborne contamination, (3) shielding and remote operations to decrease radiation exposure, and (4) equipment and ventilation design to provide flexibility and improved process performance.

Perkins, W.C.; Lee, R.; Allen, P.M.; Gouge, A.P.

1991-07-01T23:59:59.000Z

114

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

Science Conference Proceedings (OSTI)

Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (Quality Assurance for Safety-Related Software) identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls that prevent or mitigate potential accidents. The development and maintenance of a collection, or ''toolbox,'' of multiple-site use, standard solution, Software Quality Assurance (SQA)-compliant safety software is one of the major improvements identified in the associated DOE Implementation Plan (IP). The DOE safety analysis toolbox will contain a set of appropriately quality-assured, configuration-controlled, safety analysis codes, recognized for DOE-broad, safety basis applications. Currently, six widely applied safety analysis computer codes have been designated for toolbox consideration. While the toolbox concept considerably reduces SQA burdens among DOE users of these codes, many users of unique, single-purpose, or single-site software may still have sufficient technical justification to continue use of their computer code of choice, but are thwarted by the multiple-site condition on toolbox candidate software. The process discussed here provides a roadmap for an equivalency argument, i.e., establishing satisfactory SQA credentials for single-site software that can be deemed ''toolbox-equivalent''. The process is based on the model established to meet IP Commitment 4.2.1.2: Establish SQA criteria for the safety analysis ''toolbox'' codes. Implementing criteria that establish the set of prescriptive SQA requirements are based on implementation plan/procedures from the Savannah River Site, also incorporating aspects of those from the Waste Isolation Pilot Plant (SNL component) and the Yucca Mountain Project. The major requirements are met with evidence of a software quality assurance plan, software requirements and design documentation, user's instructions, test report, a configuration and control procedure, an error notification and corrective action process, and evidence of available training on use of the software. The process is best performed with an independent SQA evaluator, i.e., a technically knowledgeable individual in the application area who is not part of the development team. The process provides a consistent, systematic approach based on the experience gained with SQA evaluations of the toolbox codes. Experience has shown that rarely will existing software be fully compliant with SQA criteria. Instead, the typical case is where SQA elements are deficient. For this case, it is recommended that supplemental remedial documentation be generated. Situations may also arise where the SQA evaluator must weigh whether the entire SQA suite be reconstituted. Regardless, the process is described sufficiently to guide a comprehensive evaluation. If the candidate software is successful in meeting process requirements, the software is ''toolbox-equivalent''. The benefit of the methodology outlined is that it provides a standard evaluation technique for choosing the most applicable software for a given application. One potential outcome is that the software of choice will be found to be applicable with ample SQA justification. Alternatively, the software in question may be found not to meet SQA process requirements. In this case, the analyst may then make an informed decision and possibly select one of the multiple-use, toolbox codes. With either outcome, the DSA is improved.

O'KULA, KR

2004-04-30T23:59:59.000Z

115

Radiation Safety and Education in the Applicants of the Final Test for the Expert of Pain Medicine  

E-Print Network (OSTI)

The C-arm fluoroscope is known as the most important equipment in pain interventions. This study was conducted to investigate the completion rate of education on radiation safety, the knowledge of radiation exposure, the use of radiation protection, and so on. Methods: Unsigned questionnaires were collected from the 27 pain physicians who applied for the final test to become an expert in pain medicine in 2011. The survey was composed of 12 questions about the position of the hospital, the kind of hospital, the use of C-arm fluoroscopy, radiation safety education, knowledge of annual permissible radiation dose, use of radiation protection, and efforts to reduce radiation exposure. Results: In this study, although most respondents (93%) had used C-arm fluoroscopy, only 33 % of the physicians completed radiation safety education. Even though nine (33%) had received education on radiation safety, none of the physicians knew the annual permissible radiation dose. In comparing the radiation safety education group and the no-education group, the rate of wearing radiation-protective glasses or goggles and the use of radiation badges or dosimeters were significantly higher in the education group. However, in the use of other protective equipment, knowledge of radiation safety, and efforts to reduce radiation exposure, there were no statistical differences between the two groups. Conclusions: The respondents knew very little about radiation safety and had low interest in their radiation exposure. To make the use of fluoroscopy safer, additional education, as well as attention to and knowledge of practices of radiation safety are required for pain physicians. (Korean J Pain 2012; 25: 16-21)

Department Of Anesthesiology; Pain Medicine; Yong Chul Kim

2011-01-01T23:59:59.000Z

116

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

Science Conference Proceedings (OSTI)

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

NONE

1995-12-01T23:59:59.000Z

117

Safety Evaluation of the FuelMaker Home Refueling Concept: Final Report  

DOE Green Energy (OSTI)

Report summarizes results of a National Renewable Energy Laboratory safety evaluation of the FuelMaker natural gas vehicle home refueling appliance (HRA, aka Phill).

Waterland, L. R.; Powars, C.; Stickles, P.

2005-02-01T23:59:59.000Z

118

The Zion integrated safety analysis for NUREG-1150  

SciTech Connect

The utility-funded Zion Probabilistic Safety Study provided not only a detailed and thorough assessment of the risk profile of Zion Unit 1, but also presented substantial advancement in the technology of probabilistic risk assessment (PRA). Since performance of that study, modifications of plant hardware, the introduction of new emergency procedures, operational experience gained, information generated by severe accident research programs and further evolution of PRA and uncertainty analysis methods have provided a basis for reevaluation of the Zion risk profile. This reevaluation is discussed in this report. 5 refs.

Unwin, S.D.; Park, C.K.

1988-01-01T23:59:59.000Z

119

Safety research plan for gas-supply technologies. Final report, March 1982-February 1983  

SciTech Connect

The objective of this study was to develop a multiyear research plan addressing the safety issues of the following gas supply technologies: conventional natural gas, including deep and sour gas wells; unconventional natural gas (Devonian shale, tight gas sands, coalbed methane, and geopressured methane); SNG from coal (surface and in situ), and SNG from biomass. A total of 51 safety issues were identified in the initial review. These safety issues were screened to eliminate those hazards which appeared to be relatively insignificant in terms of accident severity or frequency, or because the potential for resolving the problem through research was considered very low. Twenty-six remaining safety issues were prioritized, and of these, 9 were selected as priority research projects: two under conventional gas; one under unconventional natural gas; and six under SNG from coal. No safety research issues in the biomass area appear to warrant priority consideration.

Tipton, L.M.; Junkin, P.D.

1983-06-01T23:59:59.000Z

120

New enhancements to SCALE for criticality safety analysis  

Science Conference Proceedings (OSTI)

As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below.

Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States). Computational Physics and Engineering Div.

1995-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Clean air program: Liquefied natural gas safety in transit operations. Final report  

SciTech Connect

The report examines the safety issues relating to the use of Liquefied natural Gas (LNG) in transit service. The surveys consisted of: (1) extensive interviews; (2) review of recrods, procedures, and plans relating to safety; (3) examination of facilities and equipment; (4) observations of operations including fueling, maintenance, morning start-up, and revenue service; (5) measurement of methane concentrations in the air where the buses are being fueled or stored. Interviews included all job categories associated with management, operations, safety, maintenance, acquisition, and support. The surveys also included an examination of the occupational hygiene aspects of LNG use.

Friedman, D.M.; Malcosky, N.D.

1996-03-01T23:59:59.000Z

122

Documented Safety Analysis for the B695 Segment  

SciTech Connect

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.

Laycak, D

2008-09-11T23:59:59.000Z

123

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2-Criteria 2-Criteria Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.2: Software Quality Assurance Plan and Criteria for the Safety Analysis Toolbox Codes U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 November 2003 Software Quality Assurance Criteria for Safety Analysis Codes November 2003 INTENTIONALLY BLANK ii Software Quality Assurance Criteria for Safety Analysis Codes November 2003 FOREWORD This document discusses the Software Quality Assurance plan, and criteria and implementation procedures to be used to evaluate designated, safety-related computer software for the

124

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

Science Conference Proceedings (OSTI)

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.

Garvin, L.J.

1997-04-28T23:59:59.000Z

125

Lawrence Livermore National Laborotory Safety Basis Assessment Final February 11, 2011  

NLE Websites -- All DOE Office Websites (Extended Search)

Lawrence Livermore National Laboratory Lawrence Livermore National Laboratory Safety Basis Assessment INTRODUCTION This site visit report documents the collective results of the review of Lawrence Livermore National Laboratory (LLNL) safety basis processes and discusses its scope, objective, results and conclusions. Appendix A provides lists of the documents, interviews, and observations and Appendix B includes the plan for the review. This combined assessment was sponsored by the National Nuclear Safety Administration (NNSA) Livermore Site Office (LSO) and conducted jointly by staff from the Office of Health, Safety and Security (HSS) and LSO. The review was conducted in late 2010 and included site visits from November 29 - December 3, 2010 and December 13-17, 2010. Overall, the LLNL programs

126

Lawrence Livermore National Laborotory Safety Basis Assessment Final February 11, 2011  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Lawrence Livermore National Laboratory Lawrence Livermore National Laboratory Safety Basis Assessment INTRODUCTION This site visit report documents the collective results of the review of Lawrence Livermore National Laboratory (LLNL) safety basis processes and discusses its scope, objective, results and conclusions. Appendix A provides lists of the documents, interviews, and observations and Appendix B includes the plan for the review. This combined assessment was sponsored by the National Nuclear Safety Administration (NNSA) Livermore Site Office (LSO) and conducted jointly by staff from the Office of Health, Safety and Security (HSS) and LSO. The review was conducted in late 2010 and included site visits from November 29 - December 3, 2010 and December 13-17, 2010. Overall, the LLNL programs

127

Lawrence Livermore Site Office Safety Basis Self-Assessment Final February 11, 2011  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Livermore Site Office Livermore Site Office Safety Basis Self-Assessment INTRODUCTION This site visit report documents the collective results of the Office of Health, Safety and Security's (HSS) assessment of National Nuclear Safety Administration (NNSA) Livermore Site Office (LSO) safety basis processes and discusses its scope, objective, results and conclusions. Appendix A provides lists of the documents, interviews, and observations and Appendix B includes the plan for the review. The assessment was sponsored by LSO as a self-assessment and conducted jointly by HSS and LSO staff. It was completed in late 2010 and included site visits from November 29 - December 3, 2010 and December 13-17, 2010. The assessment revealed that LSO has implemented appropriate plans, procedures, and

128

Lawrence Livermore Site Office Safety Basis Self-Assessment Final February 11, 2011  

NLE Websites -- All DOE Office Websites (Extended Search)

Livermore Site Office Livermore Site Office Safety Basis Self-Assessment INTRODUCTION This site visit report documents the collective results of the Office of Health, Safety and Security's (HSS) assessment of National Nuclear Safety Administration (NNSA) Livermore Site Office (LSO) safety basis processes and discusses its scope, objective, results and conclusions. Appendix A provides lists of the documents, interviews, and observations and Appendix B includes the plan for the review. The assessment was sponsored by LSO as a self-assessment and conducted jointly by HSS and LSO staff. It was completed in late 2010 and included site visits from November 29 - December 3, 2010 and December 13-17, 2010. The assessment revealed that LSO has implemented appropriate plans, procedures, and

129

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

SciTech Connect

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

Camper, L.W.; Schlueter, J.; Woods, S. [and others

1997-05-01T23:59:59.000Z

130

FINAL REPORT ANALYSIS of KANSAS CITY, MARMATON, and CHEROKEE GROUP  

E-Print Network (OSTI)

FINAL REPORT ANALYSIS of KANSAS CITY, MARMATON, and CHEROKEE GROUP COAL and SHALE SAMPLES for GAS gas contents: unit, depth (desorbed gas) · Eudora Shale Member, 242' 0" to 242' 8.5" (no significant gas) · Muncie Creek Shale Member, 438' 4.5" to 439' 4" (no significant gas) · Quivira Shale Member

Peterson, Blake R.

131

Simplifying documentation while approaching site closure: integrated health & safety plans as documented safety analysis  

Science Conference Proceedings (OSTI)

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D&D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D&D I-HASP as an example.

Brown, Tulanda

2003-06-01T23:59:59.000Z

132

Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement  

Science Conference Proceedings (OSTI)

In October 2002, DOE issued the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE 2002) that provided an analysis of the potential environmental consequences of alternatives/options for the management and disposition of Sodium Bearing Waste (SBW), High-Level Waste (HL W) calcine, and HLW facilities at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL), now known as the Idaho National Laboratory (INL) and referred to hereafter as the Idaho Site. Subsequent to the issuance of the Final EIS, DOE included the requirement for treatment of SBW in the Request for Proposals for Environmental Management activities on the Idaho Site. The new Idaho Cleanup Project (ICP) Contractor identified Steam Reforming as their proposed method to treat SBW; a method analyzed in the Final EIS as an option to treat SBW. The proposed Steam Reforming process for SBW is the same as in the Final EIS for retrieval, treatment process, waste form and transportation for disposal. In addition, DOE has updated the characterization data for both the HLW Calcine (BBWI 2005a) and SBW (BBWI 2004 and BBWI 2005b) and identified two areas where new calculation methods are being used to determine health and safety impacts. Because of those changes, DOE has prepared this supplement analysis to determine whether there are ''substantial changes in the proposed action that are relevant to environmental concerns'' or ''significant new circumstances or information'' within the meaning of the Council of Environmental Quality and DOE National Environmental Policy Act (NEPA) Regulations (40 CFR 1502.9 (c) and 10 CFR 1021.314) that would require preparation of a Supplemental EIS. Specifically, this analysis is intended to determine if: (1) the Steam Reforming Option identified in the Final EIS adequately bounds impacts from the Steam Reforming Process proposed by the new ICP Contractor using the new characterization data, (2) the new characterization data is significantly different than the data presented in the Final EIS, (3) the new calculation methods present a significant change to the impacts described in the Final EIS, and (4) would the updated characterization data cause significant changes in the environmental impacts for the action alternatives/options presented in the Final EIS. There are no other aspects of the Final EIS that require additional review because DOE has not identified any additional new significant circumstances or information that would warrant such a review.

N /A

2005-06-30T23:59:59.000Z

133

Documented Safety Analysis for the Waste Storage Facilities March 2010  

SciTech Connect

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).

Laycak, D T

2010-03-05T23:59:59.000Z

134

Phase 2 safety analysis report: National Synchrotron Light Source  

SciTech Connect

The Phase II program was established in order to provide additional space for experiments, and also staging and equipment storage areas. It also provides additional office space and new types of advanced instrumentation for users. This document will deal with the new safety issues resulting from this extensive expansion program, and should be used as a supplement to BNL Report No. 51584 ''National Synchrotron Light Source Safety Analysis Report,'' July 1982 (hereafter referred to as the Phase I SAR). The initial NSLS facility is described in the Phase I SAR. It comprises two electron storage rings, an injection system common to both, experimental beam lines and equipment, and office and support areas, all of which are housed in a 74,000 sq. ft. building. The X-ray Ring provides for 28 primary beam ports and the VUV Ring, 16. Each port is capable of division into 2 or 3 separate beam lines. All ports receive their synchrotron light from conventional bending magnet sources, the magnets being part of the storage ring lattice. 4 refs.

Stefan, P. (ed.)

1989-06-01T23:59:59.000Z

135

Documented Safety Analysis for the Waste Storage Facilities  

Science Conference Proceedings (OSTI)

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).

Laycak, D

2008-06-16T23:59:59.000Z

136

Issues related to criticality safety analysis for burnup credit applications  

SciTech Connect

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.

DeHart, M.D.; Parks, C.V.

1995-12-01T23:59:59.000Z

137

Safety-related requirements for photovoltaic modules and arrays. Final report  

DOE Green Energy (OSTI)

Underwriters Laboratories has conducted a study to identify and develop safety requirements for photovoltaic module and panel designs and configurations for residential, intermediate, and large scale applications. Concepts for safety systems, where each system is a collection of subsystems which together address the total anticipated hazard situation, are described. Descriptions of hardware, and system usefulness and viability are included. This discussion of safety systems recognizes that there is little history on which to base the expected safety related performance of a photovoltaic system. A comparison of these systems, as against the provisions of the 1984 National Electrical Code covering photovoltaic systems is made. A discussion of the UL investigation of the photovoltaic module evaluated to the provisions of the Proposed UL Standard for Flat-Plate Photovoltaic Modules and Panels is included. Grounding systems, their basis and nature, and the advantages and disadvantages of each are described. The meaning of frame grounding, circuit grounding, and the type of circuit ground are covered. The development of the Standard for Flat-Plate Photovoltaic Modules and Panels has continued, and with both industry comment and a product submittal and listing, the Standard has been refined to a viable document allowing an objective safety review of photovoltaic modules and panels. How this document, and other UL documents would cover investigations of certain other photovoltaic system components is described.

Levins, A.

1984-03-01T23:59:59.000Z

138

Safety analysis of the 700-horsepower combustion test facility  

SciTech Connect

The objective of the program reported herein was to provide a Safety Analysis of the 700 h.p. Combustion Test Facility located in Building 93 at the Pittsburgh Energy Technology Center. Extensive safety related measures have been incorporated into the design, construction, and operation of the Combustion Test Facility. These include: nitrogen addition to the coal storage bin, slurry hopper, roller mill and pulverizer baghouse, use of low oxygen content combustion gas for coal conveying, an oxygen analyzer for the combustion gas, insulation on hot surfaces, proper classification of electrical equipment, process monitoring instrumentation and a planned remote television monitoring system. Analysis of the system considering these factors has resulted in the determination of overall probabilities of occurrence of hazards as shown in Table I. Implementation of the recommendations in this report will reduce these probabilities as indicated. The identified hazards include coal dust ignition by hot ductwork and equipment, loss of inerting within the coal conveying system leading to a coal dust fire, and ignition of hydrocarbon vapors or spilled oil, or slurry. The possibility of self-heating of coal was investigated. Implementation of the recommendations in this report will reduce the ignition probability to no more than 1 x 10/sup -6/ per event. In addition to fire and explosion hazards, there are potential exposures to materials which have been identified as hazardous to personal health, such as carbon monoxide, coal dust, hydrocarbon vapors, and oxygen deficient atmosphere, but past monitoring experience has not revealed any problem areas. The major environmental hazard is an oil spill. The facility has a comprehensive spill control plan.

Berkey, B.D.

1981-05-01T23:59:59.000Z

139

Performance and Safety Analysis of a Generic Small Modular Reactor  

E-Print Network (OSTI)

The high and ever growing demand for electricity coupled with environmental concerns and a worldwide desire to shed petroleum dependence, all point to a shift to utilization of renewable sources of energy. The under developed nature of truly renewable energy sources such as, wind and solar, along with their limitations on the areas of applicability and the energy output calls for a renaissance in nuclear energy. In this second nuclear era, deliberately small reactors are poised to play a major role with a number of Small Modular Reactors (SMRs) currently under development in the U.S. In this work, an SMR model of the Integral Pressurized Water Reactor (IPWR) type is created, analyzed and optimized to meet the publically available performance criteria of the mPower SMR from B&W. The Monte Carlo codes MCNP5/MCNPX are used to model the core. Fuel enrichment, core inventory, core size are all variables optimized to meet the set goals of core lifetime and fuel utilization (burnup). Vital core behavior characteristics such as delayed neutron fraction and reactivity coefficients are calculated and shown to be typical of larger PWR systems, which is necessary to ensure the inherent safety and to achieve rapid deployment of the reactor by leveraging the vast body of operational experience amassed with the larger commercial PWRs. Inherent safety of the model is analyzed with the results of an analytical single channel analysis showing promising behavior in terms of axial and radial fuel element temperature distributions, the critical heat flux, and the departure from nucleate boiling ratio. The new fleet of proposed SMRs is intended to have increased proliferation resistance (PR) compared to the existing fleet of operating commercial PWRs. To quantify this PR gain, a PR analysis is performed using the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR) code developed by the Nuclear Science and Security Policy Institute at Texas A&M University. The PRAETOR code uses multi-attribute utility analysis to combine 63 factors affecting the PR value of a facility into a single metric which is easily comparable. The analysis compared hypothetical spent fuel storage facilities for the SMR model spent fuel assembly and one for spent fuel from a Westinghouse AP1000. The results showed that from a fuel material standpoint, the SMR and AP1000 had effectively the same PR value. Unable to analyze security systems and methods employed at specific nuclear power plant sites, it is premature to conclude that the SMR plants will not indeed show increased PR as intended.

Kitcher, Evans Damenortey, 1987-

2012-12-01T23:59:59.000Z

140

Safety analysis of natural gas vehicles transiting highway tunnel  

Science Conference Proceedings (OSTI)

A safety analysis was performed to assess the relative hazard of compressed natural gas (CNG) fueled vehicles traveling on various tunnels and bridges in New York City. The study considered those hazards arising from the release of fuel from CNG vehicles ranging in size from a passenger sedan to a full size 53 passenger bus. The approach used was to compare the fuel hazard of CNG vehicles to the fuel hazard of gasoline vehicles. The risk was assessed by estimating the frequency of occurrence and the severity of the hazard. The methodology was a combination of analyzing accident data, performing a diffusion analysis of the gas released in the tunnel and determining the consequences of ignition. Diffusion analysis was performed using the TEMPEST code for various accident scenarios resulting in CNG release inside the Holland Tunnel. The study concluded that the overall hazard of CNG vehicles transiting a ventilated tunnel is less than the hazard from a comparable gasoline fueled vehicle. 134 refs., 23 figs., 24 tabs.

Shaaban, S.H.; Zuzovsky, M.; Anigstein, R.

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Fluor Daniel Hanford Inc. integrated safety management system phase 1 verification final report  

SciTech Connect

The purpose of this review is to verify the adequacy of documentation as submitted to the Approval Authority by Fluor Daniel Hanford, Inc. (FDH). This review is not only a review of the Integrated Safety Management System (ISMS) System Description documentation, but is also a review of the procedures, policies, and manuals of practice used to implement safety management in an environment of organizational restructuring. The FDH ISMS should support the Hanford Strategic Plan (DOE-RL 1996) to safely clean up and manage the site's legacy waste; deploy science and technology while incorporating the ISMS theme to ''Do work safely''; and protect human health and the environment.

PARSONS, J.E.

1999-10-28T23:59:59.000Z

142

Hanford Waste Vitrification Plant project Preliminary Safety Analysis Report comment response records  

SciTech Connect

The initial draft version of the Hanford Waste Vitrification Plant (HWVP) Preliminary Safety Analysis Report (PSAR) was issued for review and comment on July 17, 1989, and was designated as Revision A. Following resolution of comments arising from the Revision A review, the PSAR was revised as a final draft. This version, which was designated as HWVP PSAR, Revision B, was issued for review and comment on July 16, 1990. The PSAR was again revised as Revision O following the resolution of the Revision B comments. The HWVP PSAR, Revision O, was approved by Westinghouse Hanford Company (Westinghouse Hanford) on May 21, 1991, and issued as WHC-EP-0250, Hanford Waste Vitrification Plant Preliminary Safety Analysis Report to the Department of Energy-Richland Operations Office (DOE-RL). While Westinghouse Hanford organizations were reviewing the drafts of the HWVP PSAR (i.e., Revisions A and B), a parallel review was being performed by the DOE-RL on both versions. This supporting document provides a summary of the PSAR Comment Response Databases for the Westinghouse Hanford and DOE-RL reviews of HWVP PSAR, Revisions A and B. This document also provides copies of all the closed-out Review Comment Records (RCR) submitted by these organizations. 8 refs., 1 fig., 2 tabs.

Herborn, D.I.; Campbell, L.M.

1991-06-01T23:59:59.000Z

143

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation  

Science Conference Proceedings (OSTI)

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.

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

2000-03-23T23:59:59.000Z

144

Analysis of the Relationship Between Vehicle Weight/Size and Safety, and Implications for Federal Fuel Economy Regulation  

E-Print Network (OSTI)

for Federal Fuel Economy Regulation Final Report preparedand have higher fuel economy, and safer than conventionaland have higher fuel economy, without sacrificing safety. 1.

Wenzel, Thomas P.

2010-01-01T23:59:59.000Z

145

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

Science Conference Proceedings (OSTI)

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

Slavich, Antoinette; Daust, James E.

1999-10-01T23:59:59.000Z

146

Geographically Based Hydrogen Consumer Demand and Infrastructure Analysis: Final Report  

NLE Websites -- All DOE Office Websites (Extended Search)

Geographically Based Hydrogen Geographically Based Hydrogen Consumer Demand and Infrastructure Analysis Final Report M. Melendez and A. Milbrandt Technical Report NREL/TP-540-40373 October 2006 NREL is operated by Midwest Research Institute ● Battelle Contract No. DE-AC36-99-GO10337 Geographically Based Hydrogen Consumer Demand and Infrastructure Analysis Final Report M. Melendez and A. Milbrandt Prepared under Task No. HF65.8310 Technical Report NREL/TP-540-40373 October 2006 National Renewable Energy Laboratory 1617 Cole Boulevard, Golden, Colorado 80401-3393 303-275-3000 * www.nrel.gov Operated for the U.S. Department of Energy Office of Energy Efficiency and Renewable Energy by Midwest Research Institute * Battelle Contract No. DE-AC36-99-GO10337 NOTICE This report was prepared as an account of work sponsored by an agency of the United States government.

147

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

STD-1083-2009 STD-1083-2009 June 2009 DOE STANDARD PROCESSING EXEMPTIONS TO NUCLEAR SAFETY RULES AND APPROVAL OF ALTERNATIVE METHODS FOR DOCUMENTED SAFETY ANALYSES U.S. Department of Energy AREA SAFT Washington D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. NOT MEASUREMENT SENSITIVE This document is available on the Department of Energy Technical Standards Program Web Page at http://www.hss.energy.gov/nuclearsafety/techstds DOE-STD-1083-2009 iii FOREWORD 1. This Department of Energy (DOE) Standard has been prepared by the Office of Quality Assurance Policy and Assistance to provide acceptable processes for: a. requesting and granting exemptions to DOE nuclear safety rules and b. requesting and approving alternate methodologies for documented safety analyses

148

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation  

SciTech Connect

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.

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

2009-11-01T23:59:59.000Z

149

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation  

SciTech Connect

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.

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

2009-11-01T23:59:59.000Z

150

LNG storage: Safety analysis. Annual report, January-December 1980  

SciTech Connect

Progress is summarized on three projects in the areas of LNG safety: Rollover phenomena; Simultaneous boiling and spreading of cryogenic liquids; Modelling of LNG tank dynamics.

Reid, R.C.; Smith, K.A.; Virk, P.S.

1981-02-01T23:59:59.000Z

151

Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings  

SciTech Connect

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.

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

2007-04-12T23:59:59.000Z

152

Optimization of Lyapunov invariants in analysis and implementation of safety-critical software systems  

E-Print Network (OSTI)

This dissertation contributes to two major research areas in safety-critical software systems, namely, software analysis, and software implementation. In reference to the software analysis problem, the main contribution ...

Roozbehani, Mardavij

2008-01-01T23:59:59.000Z

153

A risk-informed approach to safety margins analysis  

SciTech Connect

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.

Curtis Smith; Diego Mandelli

2013-07-01T23:59:59.000Z

154

Development of an auditable safety analysis in support of a radiological facility classification  

SciTech Connect

In recent years, U.S. Department of Energy (DOE) facilities commonly have been classified as reactor, non-reactor nuclear, or nuclear facilities. Safety analysis documentation was prepared for these facilities, with few exceptions, using the requirements in either DOE Order 5481.1B, Safety Analysis and Review System; or DOE Order 5480.23, Nuclear Safety Analysis Reports. Traditionally, this has been accomplished by development of an extensive Safety Analysis Report (SAR), which identifies hazards, assesses risks of facility operation, describes and analyzes adequacy of measures taken to control hazards, and evaluates potential accidents and their associated risks. This process is complicated by analysis of secondary hazards and adequacy of backup (redundant) systems. The traditional SAR process is advantageous for DOE facilities with appreciable hazards or operational risks. SAR preparation for a low-risk facility or process can be cost-prohibitive and quite challenging because conventional safety analysis protocols may not readily be applied to a low-risk facility. The DOE Office of Environmental Restoration and Waste Management recognized this potential disadvantage and issued an EM limited technical standard, No. 5502-94, Hazard Baseline Documentation. This standard can be used for developing documentation for a facility classified as radiological, including preparation of an auditable (defensible) safety analysis. In support of the radiological facility classification process, the Uranium Mill Tailings Remedial Action (UMTRA) Project has developed an auditable safety analysis document based upon the postulation criteria and hazards analysis techniques defined in DOE Order 5480.23.

Kinney, M.D. [Roy F. Weston, Inc., Rockville, MD (United States); Young, B. [Dept. of Energy, Albuquerque, NM (United States)

1995-03-01T23:59:59.000Z

155

Development of a multichannel analysis code for the MITR-III safety analysis  

SciTech Connect

This paper describes the development of a MULti-CHannel analysis (MULCH-II) code to be used for the safety analysis of the Massachusetts Institute of Technology Research Reactor (MITR). The code models the primary and the secondary coolant systems with special emphasis on analysis of detailed thermal-hydraulic conditions in the core region. The hot channel is modeled in parallel with the average channels to predict conditions in the core during a flow excursion instability. Fuel and cladding temperatures are calculated under all conditions so that the margin to fuel failure is given in addition to the thermal-hydraulic conditions.

Hu, Lin-Wen; Bernard, J.A. [Massachusetts Inst. of Technology, Cambridge, MA (United States)

1996-12-31T23:59:59.000Z

156

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

Science Conference Proceedings (OSTI)

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.

NONE

1995-09-01T23:59:59.000Z

157

Critical review of the reactor-safety study radiological health effects model. Final report  

Science Conference Proceedings (OSTI)

This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

1983-03-01T23:59:59.000Z

158

Alternative fueled vehicle fleet safety experience. Final report, September 1994-March 1995  

SciTech Connect

The study was initiated to gather information on the safety performance of alternative fueled vehicles from fleet operators experienced in the day to day operation of these vehicles. Eight fleets and one compressed natural gas (CNG) vehicle converter were visitied during the course of the study. The types of fleets visited consisted of these with vehicles fueled with CNG, liquefied natural gas (LNG), liquefied petroleum gas (LPG), and electric vehicles (EVs). Three CNG fleets, two LNG fleets, one EV fleet, and two LPG fleets were visitied in addition to one CNG converter. Items discussed with the fleet operators included fuel system performance in the crash environment as well as safety related problems encountered during the refueling operation and when maintaining the vehicles. The fleets visited have experienced no accidents where the fuel system has been jeopardized and no injury to personnel that can be attributed to the alternative fuel system. However, the accident experience of the fleets visited is very limited. Many of the problems with alternative fueled vehicles experienced in the past have been corrected by advances in the state of the art and improvements in system components. Improvements continue to be made.

Morris, J.B.

1995-03-01T23:59:59.000Z

159

Portsmouth DUF6 Conversion Final EIS - Chapter 6: Environmental and Occupational Safety and Health Permits and Compliance Requirements  

NLE Websites -- All DOE Office Websites (Extended Search)

Portsmouth DUF Portsmouth DUF 6 Conversion Final EIS 6 ENVIRONMENTAL AND OCCUPATIONAL SAFETY AND HEALTH PERMITS AND COMPLIANCE REQUIREMENTS 6.1 DUF 6 CYLINDER MANAGEMENT AND CONSTRUCTION AND OPERATION OF A DUF 6 CONVERSION FACILITY DUF 6 cylinder management as well as construction and operation of the proposed DUF 6 conversion facility would be subject to many federal, state, and local requirements. In accordance with such legal requirements, a variety of permits, licenses, and other consents must be obtained. Table 6.1 at the end of this chapter lists those that may be needed. The status of each is indicated on the basis of currently available information. However, because the DUF 6 project is still at an early stage, the information in Table 6.1 should not be considered comprehensive or

160

Consolidation of public safety wireless networks: An options-based economic analysis of numerous scenarios  

Science Conference Proceedings (OSTI)

The Korean National Emergency Management Agency proposed to replace existing public safety wireless networks of 46 agencies with a nation-wide consolidated network. This study compares the public-private partnership alternative of sharing a network with ... Keywords: Consolidation, Exploratory modeling, Feasibility study, Public safety wireless network, Public-private partnership, Real option analysis

Sungho Lee

2011-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

An impact analysis method for safety-critical user interface design  

Science Conference Proceedings (OSTI)

We describe a method of assessing the implications for human error on user interface design of safety-critical systems. In previous work we have proposed a taxonomy of influencing factors that contribute to error. In this article, components of the taxonomy ... Keywords: Bayesian belief networks, human error, safety-critical, scenario-based casual analysis

Julia Galliers; Alistair Sutcliffe; Shailey Minocha

1999-12-01T23:59:59.000Z

162

Final Report K I N E SAFETY EVALUATION PROJECT RULIS ON  

Office of Legacy Management (LM)

K K I N E SAFETY EVALUATION PROJECT RULIS ON By ,R. L . Bolmer U . S . Bureau of Mines Denver ,Mining Research Center ' Denver, Colorado January 1 0 , 1970 DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. CONTENTS ! P a g e Summary . . . . . . . . . . . . . . . . . . . . . . . . . 1 I n t r o d u c t i o n H i s t o r i c a l d e s c r i p t i o n . . . . . . . . . . . . . . . . 2 - Mine S a f e t y E v a l u a t i o n Program . . . . . . . . . . . . 3 G e n e r a l s e t t i n g . . . . . . . . . . . . . . . . . . . . . 3 Mines i n p r o j e c t a r e a . . . . . . . . . . . . . . . . . . 4 Mine e f f e c t s s a f e t y e v a l u a t i o n Mine e v a c u a t i o n . . . . . . . . . . . . . . . . . . . . 6 P r e - and p o s t - s h o t mine i n s p e c t i o n s . . . . . . . . . . 7 . . . . . . . . . . . . . . . . Mine s t r u c t u r a l damage 8 Cameo mine. . . . . . . . . . . . . . . . . . . . . 9 . . . . . . . . . . . . . . . . . . Red Canon mine. 10

163

A final report on the Great Plains Gasification Project's environmental, health, and safety information data system  

Science Conference Proceedings (OSTI)

This report describes Oak Ridge National Laboratory's (ORNLs) role in providing information to Department of Energy (DOE) on environmental data generated at the Great Plains Coal Gasification Project (GPCGP) in Beulah, North Dakota. An information system, the Fossil Energy (FE) Environmental, Health, and Safety Information System (EHSIS) was developed at ORNL to assist in tracking, analyzing, and making readily available significant environmental information derived from Great Plains. The Great Plains module with its numerous files (e.g., Gasification Bibliography, Gasification Tables, and Great Plains Gasification Project -- Permits, Standards, or Exceedences/Incidents) is a major technical area located within the information system. Over 1388 Great Plains documents have been reviewed, abstracted, and made available on-line in the information system. Also in the information system are 911 tables of selected environmental data including monitoring data from the following six subject areas: (1) air quality; (2) water quality; (3) solid wastes; (4) hazardous wastes; (5) industrial hygiene; and (6) surface mining. 14 refs., 4 figs.

Noghrei-Nikbakht, P.A.; Roseberry, L.M.

1989-12-01T23:59:59.000Z

164

Electrical Safety - Monthly Analyses of Electrical Safety Occurrences  

NLE Websites -- All DOE Office Websites (Extended Search)

Office of Analysis Office of Analysis Operating Experience Committee Safety Alerts Safety Bulletins Annual Reports Special Operations Reports Safety Advisories Special Reports Causal Analysis Reviews Contact Us HSS Logo Electrical Safety Monthly Analyses of Electrical Safety Occurrences 2013 September 2013 Electrical Safety Occurrences August 2013 Electrical Safety Occurrences July 2013 Electrical Safety Occurrences June 2013 Electrical Safety Occurrences May 2013 Electrical Safety Occurrences April 2013 Electrical Safety Occurrences March Electrical Safety Occurrence February Electrical Safety Occurrence January Electrical Safety Occurrence 2012 December Electrical Safety Occurrence November Electrical Safety Occurrence October Electrical Safety Occurrence September Electrical Safety Occurrence

165

Knowledge-centric and language independent framework for safety analysis tools  

Science Conference Proceedings (OSTI)

This paper presents a knowledge-centric and language independent framework and its application to develop safety analysis tools for avionics systems. A knowledge-centric approach is important to address domain-specific needs, with respect to the types ...

S. C. Kothari; Luke Bishop; Jeremias Sauceda; Gary Daugherty

2004-03-01T23:59:59.000Z

166

FINAL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2 2 FINAL ENVIRONMENTAL ASSESSMENT FOR EXIDE TECHNOLOGIES ELECTRIC DRIVE VEHICLE BATTERY AND COMPONENT MANUFACTURING INITIATIVE APPLICATION, BRISTOL, TN, AND COLUMBUS, GA U.S. Department of Energy National Energy Technology Laboratory March 2010 DOE/EA-1712 FINAL ENVIRONMENTAL ASSESSMENT FOR EXIDE TECHNOLOGIES ELECTRIC DRIVE VEHICLE BATTERY AND COMPONENT MANUFACTURING INITIATIVE APPLICATION, BRISTOL, TN, AND COLUMBUS, GA U.S. Department of Energy National Energy Technology Laboratory March 2010 DOE/EA-1712 iii COVER SHEET Responsible Agency: U.S. Department of Energy (DOE) Title: Environmental Assessment for Exide Technologies Electric Drive Vehicle Battery and Component Manufacturing Initiative Application, Bristol, TN, and Columbus, GA

167

Savannah River Site high-level waste safety issues: The need for final disposal of the wastes  

DOE Green Energy (OSTI)

Using new criteria developed by the High-Level Waste Tank Safety Task Force, the Savannah River Site (SRS) identified six safety issues in the SRS tank farms. None of the safety issues were priority 1, the most significant issues handled by the Task Force. This paper discusses the safety issues and the programs for resolving each of them.

d`Entremont, P.D.; Hobbs, D.T.

1991-12-31T23:59:59.000Z

168

Savannah River Site high-level waste safety issues: The need for final disposal of the wastes  

DOE Green Energy (OSTI)

Using new criteria developed by the High-Level Waste Tank Safety Task Force, the Savannah River Site (SRS) identified six safety issues in the SRS tank farms. None of the safety issues were priority 1, the most significant issues handled by the Task Force. This paper discusses the safety issues and the programs for resolving each of them.

d'Entremont, P.D.; Hobbs, D.T.

1991-01-01T23:59:59.000Z

169

Thread-safety in an MPI implementation: Requirements and analysis  

Science Conference Proceedings (OSTI)

The MPI-2 Standard has carefully specified the interaction between MPI and user-created threads. The goal of this specification is to allow users to write multithreaded MPI programs while also allowing MPI implementations to deliver high performance. ... Keywords: MPI implementation, Message-passing interface (MPI), Multithreaded programming, Thread-safety

William Gropp; Rajeev Thakur

2007-09-01T23:59:59.000Z

170

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

SciTech Connect

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.

Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

2011-06-01T23:59:59.000Z

171

Hogged Wood Fuel Supply and Price Analysis : Final Report.  

SciTech Connect

This study discusses the factors that determine the supply and demand for hogged wood in the Pacific Northwest, with particular emphasis on the role of the regional pulp and paper industry and lumber industry. Because hogged wood is often a substitute for conventional fuels, the consumption and price of natural gas, electricity, fuel oil and coal are also addressed. A detailed and comprehensive examination of the indicies relating to the hogged wood market is provided, including analysis and graphing of all time series variables. A spreadsheet- based forecasting model is developed and presented with an emphasis on explaining the process used to arrive at the final model. 42 refs., 46 figs., 14 tabs. (MHB)

Biederman, Richard T.; Blazek, Christopher F.

1991-05-01T23:59:59.000Z

172

Urban Integrated Industrial Cogeneration Systems Analysis. Phase II final report  

SciTech Connect

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

Not Available

1984-01-01T23:59:59.000Z

173

Coordination Meeting with National Nuclear Security Administration Los Alamos Field Office Safety Basis Review Team Leader for Transuranic Waste Facility Preiminary Documented Safety Analysis Report  

NLE Websites -- All DOE Office Websites (Extended Search)

HIAR-LANL-2013-04-08 HIAR-LANL-2013-04-08 Site: Los Alamos National Laboratory Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Coordination Meeting with National Nuclear Security Administration Los Alamos Field Office Safety Basis Review Team Leader for Transuranic Waste Facility Preliminary Documented Safety Analysis Report Dates of Activity : 04/08/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff visited the Los Alamos National Laboratory (LANL) to coordinate with the National Nuclear Security Administration (NNSA) Los Alamos Field Office (NA-00-LA) Safety Basis Review Team (SBRT) Leader for review of the revised preliminary documented safety analysis (PDSA) for the Transuranic Waste

174

Coordination Meeting with National Nuclear Security Administration Los Alamos Field Office Safety Basis Review Team Leader for Transuranic Waste Facility Preiminary Documented Safety Analysis Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

HIAR-LANL-2013-04-08 HIAR-LANL-2013-04-08 Site: Los Alamos National Laboratory Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Coordination Meeting with National Nuclear Security Administration Los Alamos Field Office Safety Basis Review Team Leader for Transuranic Waste Facility Preliminary Documented Safety Analysis Report Dates of Activity : 04/08/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS) staff visited the Los Alamos National Laboratory (LANL) to coordinate with the National Nuclear Security Administration (NNSA) Los Alamos Field Office (NA-00-LA) Safety Basis Review Team (SBRT) Leader for review of the revised preliminary documented safety analysis (PDSA) for the Transuranic Waste

175

The Independent Technical Analysis Process Final Report 2006-2007.  

DOE Green Energy (OSTI)

The Bonneville Power Administration (BPA) contracted with the Pacific Northwest National Laboratory (PNNL) to provide technical analytical support for system-wide fish passage information (BPA Project No. 2006-010-00). The goal of this project was to produce rigorous technical analysis products using independent analysts and anonymous peer reviewers. This project provided an independent technical source for non-routine fish passage analyses while allowing routine support functions to be performed by other well-qualified entities. The Independent Technical Analysis Process (ITAP) was created to provide non-routine analysis for fish and wildlife agencies and tribes in particular and the public in general on matters related to juvenile and adult salmon and steelhead passage through the mainstem hydrosystem. The process was designed to maintain the independence of analysts and reviewers from parties requesting analyses, to avoid potential bias in technical products. The objectives identified for this project were to administer a rigorous, transparent process to deliver unbiased technical assistance necessary to coordinate recommendations for storage reservoir and river operations that avoid potential conflicts between anadromous and resident fish. Seven work elements, designated by numbered categories in the Pisces project tracking system, were created to define and accomplish project goals as follows: (1) 118 Coordination - Coordinate technical analysis and review process: (a) Retain expertise for analyst/reviewer roles. (b) Draft research directives. (c) Send directive to the analyst. (d) Coordinate two independent reviews of the draft report. (e) Ensure reviewer comments are addressed within the final report. (2) 162 Analyze/Interpret Data - Implement the independent aspects of the project. (3) 122 Provide Technical Review - Implement the review process for the analysts. (4) 132 Produce Annual Report - FY06 annual progress report with Pisces Disseminate (5) 161 Disseminate Raw/Summary Data and Results - Post technical products on the ITAP web site. (6) 185-Produce Pisces Status Report - Provide periodic status reports to BPA. (7) 119 Manage and Administer Projects - project/contract administration.

Duberstein, Corey; Ham, Kenneth; Dauble, Dennis; Johnson, Gary [Pacific Northwest National Laboratory

2007-03-01T23:59:59.000Z

176

FAQS JOB TASK ANALYSIS - Electrical Systems and Safety Oversight  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Electrical Systems and Safety Oversight Electrical Systems and Safety Oversight Step 1 Identify and evaluate tasks - Develop a comprehensive list of tasks that define the job. o A great starting point is the list of Duties and Responsibilities from the FAQS. o Give careful thought to additional tasks that could be considered. o Don't worry about deleting tasks at this point - that is a part of the process further down. - List the tasks (and their sources, e.g., Duties and Responsibilities #1) in the chart below. - Discuss each task as a group and come to a consensus pertaining to Importance and Frequency of the task (i.e., each team member can consent to the assigned value, even if they don't exactly agree with it). - When all values have been assigned, consider as a group deleting tasks that receive

177

Final  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

, , Final for Vegetation Control at VHF Stations, Microwave Stations, Electrical Substations, and Pole Yards . Environmental Assessment Prepared for Southwestern Power Administration U.S. Department of Energy - _ . . . " Prepared by Black & Veatch October 13,1995 ' Table of Contents 1 . 0 Purpose and Need for Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0 Description of the Alternatives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 Alternative 1 . No Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2 Alternative 2 . Mechanical and Manual Control . . . . . . . . . . . . . . . . . . . 2.3 Alternative 3 . Proposed Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3.1 Foliar Spray Application . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3.2 Soil-Spot Application . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

178

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

SciTech Connect

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

NONE

1997-05-01T23:59:59.000Z

179

Current Safety Performance Trends  

NLE Websites -- All DOE Office Websites (Extended Search)

Environmental Protection, Sustainability Support & Corporate Safety Analysis HS-20 Home Mission & Functions Office of Sustainability, Environment, Safety and Anaylsis (SESA) ...

180

Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility  

SciTech Connect

The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL`s Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed.

Charak, I; Pedersen, D.R. [Argonne National Lab., IL (United States); Forrester, R.J.; Phipps, R.D. [Argonne National Lab., Idaho Falls, ID (United States)

1993-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

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

DOE Green Energy (OSTI)

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.

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 (Sala & Associates); Bessette, Gregory Carl; Lipinski, Ronald J.

2006-09-01T23:59:59.000Z

182

Analysis of Fundamental NIST Sphere Experiments Related to Criticality Safety  

SciTech Connect

A series of neutron transport experiments was performed in 1989 and 1990 at NIST (National Institute of Standards and Technology) using a spherical stainless steel container and fission chambers. These experiments were performed to help understand errors observed in criticality calculations for arrays of individually subcritical components, particularly solution arrays [1-3]. They were supported by the U.S. Department of Energy, Environment and Health, Nuclear Criticality Technology and Safety Project. The intent was to evaluate the possibility that the criticality prediction errors stem from errors in the calculation of neutron leakage from individual components of the array. Thus, the explicit product of the experiments was the measurement of the leakage flux, as characterized by various Cd-shielded and unshielded fission rates. Because the various fission rates have different neutron-energy sensitivities, collectively they give an indication of the energy dependence of the leakage flux. Leakage and moderation were varied systematically through the use of different diameter spheres, with and without water. Some of these experiments with bare fission chambers have been evaluated by the International Criticality Safety Benchmark Evaluation Project (ICSBEP)[4].

Kim, Soon S.

2007-06-01T23:59:59.000Z

183

Advanced methods development for LWR trsansient analysis, final report : 1981-1982  

E-Print Network (OSTI)

The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid ...

Griggs, D. P.

1982-01-01T23:59:59.000Z

184

340 Waste handling Facility Hazard Categorization and Safety Analysis  

DOE Green Energy (OSTI)

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

T. J. Rodovsky

2010-10-25T23:59:59.000Z

185

Analysis of the optics of the Final Focus Test Beam using Lie algebra based techniques  

Science Conference Proceedings (OSTI)

This report discusses the analysis of the beam optics of the final focus test beam at the Stanford Linear Collider using Lie algebra. (LSP).

Roy, G.J.

1992-09-01T23:59:59.000Z

186

The Effect on Electricity Consumption of the Commonwealth Edison Customer Applications Program: Phase 2 Final Analysis  

Science Conference Proceedings (OSTI)

This report describes the final Phase 2 analysis of the effects on residential customers' energy consumption patterns of Commonwealth Edison's (ComEd's) Customer Application Program (CAP).

2011-10-20T23:59:59.000Z

187

DOE: Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities, 10j/24/03  

Energy.gov (U.S. Department of Energy (DOE))

This document contains software quality assurance (SQA) assessment criteria and guidelines for assessing the safety software currently in use in the safety analysis and design of structures,...

188

Volume II - Accident and Operational Safety Analysis Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

into analysis tools: Culture Attribute Matrix (CAM); Missed Opportunity Matrix (MOM); Human Error Precursor Matrix (TWIN); Latent Organizational Weakness Table (LOW) *...

189

Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

by industry for recycling facilities 2 , is a systematic analysis to identify facility and external hazards and their potential for initiating accident sequences, the...

190

Criticality safety analysis of Hanford Waste Tank 241-101-SY  

SciTech Connect

As part of a safety assessment for proposed pump mixing operations to mitigate episodic gas releases in Tank 241-101-SY at the Hanford Site, Richland, Washington, a criticality safety analysis was made using the Sn transport code ONEDANT. The tank contains approximately one million gallons of waste and an estimated 910 G of plutonium. the criticality analysis considers reconfiguration and underestimation of plutonium content. The results indicate that Tank SY-101 does not present a criticality hazard. These methods are also used in criticality analyses of other Hanford tanks.

Perry, R.T.; Sapir, J.L.; Krohn, B.J.

1993-12-31T23:59:59.000Z

191

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

Science Conference Proceedings (OSTI)

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.

NONE

1994-10-01T23:59:59.000Z

192

Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Executive Summary This paper addresses why the use of an Integrated Safety Analysis ("ISA") is appropriate for fuel recycling facilities 1 which would be licensed under new regulations currently being considered by NRC. The use of the ISA for fuel facilities under Part 70 is described and compared to the use of a Probabilistic Risk Assessment ("PRA") for reactor facilities. A basis is provided for concluding that future recycling facilities - which will possess characteristics similar to today's fuel cycle facilities and distinct from reactors - can best be assessed using established qualitative or semi-quantitative ISA techniques to achieve and demonstrate safety in an effective and efficient manner.

193

Fuzzy-algebra uncertainty analysis for abnormal-environment safety assessment  

Science Conference Proceedings (OSTI)

Many safety (risk) analyses depend on uncertain inputs and on mathematical models chosen from various alternatives, but give fixed results (implying no uncertainty). Conventional uncertainty analyses help, but are also based on assumptions and models, the accuracy of which may be difficult to assure. Some of the models and assumptions that on cursory examination seem reasonable can be misleading. As a result, quantitative assessments, even those accompanied by uncertainty measures, can give unwarranted impressions of accuracy. Since analysis results can be a major contributor to a safety-measure decision process, risk management depends on relating uncertainty to only the information available. The uncertainties due to abnormal environments are even more challenging than those in normal-environment safety assessments, and therefore require an even more cautious approach. A fuzzy algebra analysis is proposed in this report that has the potential to appropriately reflect the information available and portray uncertainties well, especially for abnormal environments.

Cooper, J.A.

1994-01-01T23:59:59.000Z

194

Analysis Of The Tank 6F Final Characterization Samples-2012  

Science Conference Proceedings (OSTI)

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm-243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite Sample 2, and highest concentrations for Composite Sample 3. The Hg and Mo results suggest possible measurement outliers. However, the magnitudes of the differences between the Hg 95% upper confidence limit (UCL95) results with and without the outlier and the magnitudes of the differences between the Mo UCL95 results with and without the outlier do not appear to have practical significance. It is recommended to remove the potential measurement outliers. Doing so is conservative in the sense of producing a higher UCL95 for Hg and Mo than if the potential outliers were included in the calculations. In contrast to the inorganic results, most of the radionuclides did not demonstrate heterogeneity among the three Tank 6F composite sample characterization results.

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

2012-09-27T23:59:59.000Z

195

ANALYSIS OF THE TANK 6F FINAL CHARACTERIZATION SAMPLES-2012  

Science Conference Proceedings (OSTI)

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm-243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite Sample 2, and highest concentrations for Composite Sample 3. The Hg and Mo results suggest possible measurement outliers. However, the magnitudes of the differences between the Hg 95% upper confidence limit (UCL95) results with and without the outlier and the magnitudes of the differences between the Mo UCL95 results with and without the outlier do not appear to have practical significance. It is recommended to remove the potential measurement outliers. Doing so is conservative in the sense of producing a higher UCL95 for Hg and Mo than if the potential outliers were included in the calculations. In contrast to the inorganic results, most of the radionuclides did not demonstrate heterogeneity among the three Tank 6F composite sample characterization results.

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

2012-06-28T23:59:59.000Z

196

Analysis of the Tank 6F Final Characterization Samples-2012  

SciTech Connect

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to provide sample preparation and analysis of the Tank 6F final characterization samples to determine the residual tank inventory prior to grouting. Fourteen residual Tank 6F solid samples from three areas on the floor of the tank were collected and delivered to SRNL between May and August 2011. These Tank 6F samples were homogenized and combined into three composite samples based on a proportion compositing scheme and the resulting composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 6F composite samples include bulk density and water leaching of the solids to account for water soluble components. The composite Tank 6F samples were analyzed and the data reported in triplicate. Sufficient quality assurance standards and blanks were utilized to demonstrate adequate characterization of the Tank 6F samples. The main evaluation criteria were target detection limits specified in the technical task request document. While many of the target detection limits were met for the species characterized for Tank 6F some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The isotopes whose detection limits were not met in all cases included Sn-126, Sb-126, Sb-126m, Eu-152, Cm- 243 and Cf-249. SRNL, in conjunction with the customer, reviewed all of these cases and determined that the impacts of not meeting the target detection limits were acceptable. Based on the analyses of variance (ANOVA) for the inorganic constituents of Tank 6F, all the inorganic constituents displayed heterogeneity. The inorganic results demonstrated consistent differences across the composite samples: lowest concentrations for Composite Sample 1, intermediate-valued concentrations for Composite Sample 2, and highest concentrations for Composite Sample 3. The Hg and Mo results suggest possible measurement outliers. However, the magnitudes of the differences between the Hg 95% upper confidence limit (UCL95) results with and without the outlier and the magnitudes of the differences between the Mo UCL95 results with and without the outlier do not appear to have practical significance. It is recommended to remove the potential measurement outliers. Doing so is conservative in the sense of producing a higher UCL95 for Hg and Mo than if the potential outliers were included in the calculations. In contrast to the inorganic results, most of the radionuclides did not demonstrate heterogeneity among the three Tank 6F composite sample characterization results.

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

2013-01-31T23:59:59.000Z

197

ANALYSIS OF THE TANK 5F FINAL CHARATERIZATION SAMPLES-2011  

SciTech Connect

The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the following: Al-26, Sn-126, Sb-126, Sb-126m, Eu-152 and Cf-249. SRNL, in conjunction with the plant customer, reviewed all these cases and determined that the impacts were negligible.

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

2012-01-20T23:59:59.000Z

198

Analysis Of The Tank 5F Final Characterization Samples-2011  

SciTech Connect

The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the following: Al-26, Sn-126, Sb-126, Sb-126m, Eu-152 and Cf-249. SRNL, in conjunction with the plant customer, reviewed all these cases and determined that the impacts were negligible.

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

2012-09-27T23:59:59.000Z

199

ANALYSIS OF THE TANK 5F FINAL CHARACTERIZATION SAMPLES-2011  

SciTech Connect

The Savannah River National Laboratory (SRNL) was requested by SRR to provide sample preparation and analysis of the Tank 5F final characterization samples to determine the residual tank inventory prior to grouting. Two types of samples were collected and delivered to SRNL: floor samples across the tank and subsurface samples from mounds near risers 1 and 5 of Tank 5F. These samples were taken from Tank 5F between January and March 2011. These samples from individual locations in the tank (nine floor samples and six mound Tank 5F samples) were each homogenized and combined in a given proportion into 3 distinct composite samples to mimic the average composition in the entire tank. These Tank 5F composite samples were analyzed for radiological, chemical and elemental components. Additional measurements performed on the Tank 5F composite samples include bulk density and water leaching of the solids to account for water soluble species. With analyses for certain challenging radionuclides as the exception, all composite Tank 5F samples were analyzed and reported in triplicate. The target detection limits for isotopes analyzed were based on customer desired detection limits as specified in the technical task request documents. SRNL developed new methodologies to meet these target detection limits and provide data for the extensive suite of components. While many of the target detection limits were met for the species characterized for Tank 5F, as specified in the technical task request, some were not met. In a few cases, the relatively high levels of radioactive species of the same element or a chemically similar element precluded the ability to measure some isotopes to low levels. The Technical Task Request allows that while the analyses of these isotopes is needed, meeting the detection limits for these isotopes is a lower priority than meeting detection limits for the other specified isotopes. The isotopes whose detection limits were not met in all cases included the following: Al-26, Sn-126, Sb-126, Sb-126m, Eu-152 and Cf-249. SRNL, in conjunction with the plant customer, reviewed all these cases and determined that the impacts were negligible.

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

2012-08-03T23:59:59.000Z

200

Preliminary Accident Analysis for Construction and Operation of the Chornobyl New Safety Confinement  

Science Conference Proceedings (OSTI)

Analysis of potential exposure of personal and population during construction and exploitation of the New Safe Confinement was made. Scenarios of hazard event development were ranked. It is shown, that as a whole construction and exploitation of the NSC are in accordance with actual radiation safety norms of Ukraine.

Batiy, Valeriy; Rubezhansky, Yruiy; Rudko, Vladimir; shcherbin, vladimir; Yegorov, V; Schmieman, Eric A.; Timmins, Douglas C.

2005-08-08T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Top-Off Safety Analysis for NSLS-II  

Science Conference Proceedings (OSTI)

Top-off injection will be adopted in NSLS-II. To ensure no injected beam can pass into experimental beamlines with open photon shutters during top-off injection, simulation studies for possible machine fault scenarios are required. We compare two available simulation methods, backward (H. Nishimura-LBL) and forward tracking (A. Terebilo-SLAC). We also discuss the tracking settings, fault scenarios, apertures and interlocks considered in the analysis.

Li,Y.; Casey, B.; Heese, R.; Hseuh, H.; Job, O.; Krinsky, S.; Parker, B.; Shaftan, T.; Sharma, S.

2009-05-04T23:59:59.000Z

202

Maintaining plant safety margins  

SciTech Connect

The Final Safety Analysis Report Forms the basis of demonstrating that the plant can operate safely and meet all applicable acceptance criteria. In order to assure that this continues through each operating cycle, the safety analysis is reexamined for each reload core. Operating limits are set for each reload core to assure that safety limits and applicable acceptance criteria are not exceeded for postulated events within the design basis. These operating limits form the basis for plant operation, providing barriers on various measurable parameters. The barriers are refereed to as limiting conditions for operation (LCO). The operating limits, being influenced by many factors, can change significantly from cycle to cycle. In order to be successful in demonstrating safe operation for each reload core (with adequate operating margin), it is necessary to continue to focus on ways to maintain/improve existing safety margins. Existing safety margins are a function of the plant type (boiling water reactor/pressurized water reactor (BWR/PWR)), nuclear system supply (NSSS) vendor, operating license date, core design features, plant design features, licensing history, and analytical methods used in the safety analysis. This paper summarizes the experience at Yankee Atomic Electric Company (YAEC) in its efforts to provide adequate operating margin for the plants that it supports.

Bergeron, P.A.

1989-01-01T23:59:59.000Z

203

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

Science Conference Proceedings (OSTI)

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.

Shedrow, C.B.

1999-11-29T23:59:59.000Z

204

Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis  

SciTech Connect

This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the U.S. Department of Energy s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multi-physics simulation tools that leverage high-performance computing to reduce uncertainty in the prediction of performance and safety of advanced reactor and fuel designs. The Toolkit effort is comprised of two major components, the Fuels Product Line (FPL), which provides tools for fuel performance analysis, and the Reactor Product Line (RPL), which provides tools for reactor performance and safety analysis. This paper provides an overview of the NEAMS RPL development effort.

Pointer, William David [ORNL; Bradley, Keith S [ORNL; Fischer, Paul F [ORNL; Smith, Micheal A [ORNL; Tautges, Timothy J [ORNL; Ferencz, Robert M [ORNL; Martineau, Richard C [ORNL; Jain, Rajeev [ORNL; Obabko, Aleksandr [Argonne National Laboratory (ANL); Billings, Jay Jay [ORNL

2013-01-01T23:59:59.000Z

205

Superconducting x-ray lithography source Phase 1 (XLS) safety analysis report  

SciTech Connect

This paper discusses safety aspects associated with the superconducting x-ray lithography source. The policy, building systems safety and storage ring systems safety are specifically addressed. (LSP)

Blumberg, L. (ed.)

1990-07-01T23:59:59.000Z

206

Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR  

SciTech Connect

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

1983-08-01T23:59:59.000Z

207

60-day waste compatibility safety issue and final results for 244-TX DCRT, grab samples TX-95-1, TX-95-2, and TX-95-3  

Science Conference Proceedings (OSTI)

Three grab samples (TX-95-1, TX-95-2, and TX-95-3) were taken from tank 241- TX-244 riser 8 on November 7, 1995 and received by the 222-S Laboratory on that same day. Samples TX-95-1 and TX-95-2 were designated as supernate liquids, and sample TX-95-3 was designated as a supernate/sludge. These samples were analyzed to support the waste compatibility safety program. Accuracy and precision criteria were met for all analyses. No notifications were required based on sample results. This document provides the analysis to support the waste compatibility safety program.

Esch, R.A.

1996-01-01T23:59:59.000Z

208

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

Charges Relating to Nuclear Reactor Safety," 1976, availableissues impor tant to nuclear reactor safety. This report wasstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

209

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

Science Conference Proceedings (OSTI)

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.

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

2006-05-18T23:59:59.000Z

210

Maintaining scale as a realiable computational system for criticality safety analysis  

SciTech Connect

Accurate and reliable computational methods are essential for nuclear criticality safety analyses. The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer code system was originally developed at Oak Ridge National Laboratory (ORNL) to enable users to easily set up and perform criticality safety analyses, as well as shielding, depletion, and heat transfer analyses. Over the fifteen-year life of SCALE, the mainstay of the system has been the criticality safety analysis sequences that have featured the KENO-IV and KENO-V.A Monte Carlo codes and the XSDRNPM one-dimensional discrete-ordinates code. The criticality safety analysis sequences provide automated material and problem-dependent resonance processing for each criticality calculation. This report details configuration management which is essential because SCALE consists of more than 25 computer codes (referred to as modules) that share libraries of commonly used subroutines. Changes to a single subroutine in some cases affect almost every module in SCALE! Controlled access to program source and executables and accurate documentation of modifications are essential to maintaining SCALE as a reliable code system. The modules and subroutine libraries in SCALE are programmed by a staff of approximately ten Code Managers. The SCALE Software Coordinator maintains the SCALE system and is the only person who modifies the production source, executables, and data libraries. All modifications must be authorized by the SCALE Project Leader prior to implementation.

Bowmann, S.M.; Parks, C.V.; Martin, S.K.

1995-04-01T23:59:59.000Z

211

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

DOE Green Energy (OSTI)

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.

MCCAIN, D.J.

2000-07-26T23:59:59.000Z

212

2011 Annual Workforce Analysis and Staffing Plan Report - NNSA for Safety and Health  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Annual Workforce Analysis and Staffing Plan Report Annual Workforce Analysis and Staffing Plan Report As of December 31, 2011 Reporting Office: NNSA NA-SH Section One: Current Mission(s) of the Organization and Potential Changes The Office of the Associate Administrator for Safety and Health (NA-SH) provides mission enabling support to the NNSA Administrator, Central Technical Authority (CTA), Acquisition Executives, senior NNSA officials, program officers and site offices. NA-SH enables other NNSA organizations to fulfill NNSA missions while protecting the environment and safeguarding the safety and health of the public and the workforce. Section Two: SITE CHARACTERISTICS TABLE 1 Number of Hazard Category 1, 2, or 3 Nuclear Facilities: HC 1: 0; HC 2: 0; HC 3: 0 Number of Radiological Facilities

213

Architecture for Interlock Systems Reliability Analysis with Regard to Safety and Availability  

E-Print Network (OSTI)

In the design of interlock loops for the signal exchange in machine protection systems, the choice of the hardware architecture impacts on machine safety and availability. The reliable performance of a machine stop (leaving the machine in a safe state) in case of an emergency, is an inherent requirement. The constraints in terms of machine availability on the other hand may differ from one facility to another. Spurious machine stops, lowering machine availability, may to a certain extent be tolerated in facilities where they do not cause undue equipment wearout. In order to compare various interlock loop architectures in terms of safety and availability, the occurrence frequencies of related scenarios have been calculated in a reliability analysis, using a generic analytical model. This paper presents the results and illustrates the potential of the analysis method for supporting the choice of interlock system architectures.

Wagner, S; Schmidt, R; Zerlauth, M; Vergara-Fernandez, A

2011-01-01T23:59:59.000Z

214

Certification process of safety analysis and risk management computer codes at the Savannah River Site  

Science Conference Proceedings (OSTI)

The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States)); Toffer, H.; Lewis, C.J.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States))

1992-01-01T23:59:59.000Z

215

Certification process of safety analysis and risk management computer codes at the Savannah River Site  

Science Conference Proceedings (OSTI)

The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Toffer, H.; Lewis, C.J.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States)

1992-05-01T23:59:59.000Z

216

Photovoltaic venture analysis. Final report. Volume III. Appendices  

DOE Green Energy (OSTI)

This appendix contains a brief summary of a detailed description of alternative future energy scenarios which provide an overall backdrop for the photovoltaic venture analysis. Also included is a summary of a photovoltaic market/demand workshop, a summary of a photovoltaic supply workshop which used cross-impact analysis, and a report on photovoltaic array and system prices in 1982 and 1986. The results of a sectorial demand analysis for photovoltaic power systems used in the residential sector (single family homes), the service, commercial, and institutional sector (schools), and in the central power sector are presented. An analysis of photovoltaics in the electric utility market is given, and a report on the industrialization of photovoltaic systems is included. A DOE information memorandum regarding ''A Strategy for a Multi-Year Procurement Initiative on Photovoltaics (ACTS No. ET-002)'' is also included. (WHK)

Costello, D.; Posner, D.; Schiffel, D.; Doane, J.; Bishop, C.

1978-07-01T23:59:59.000Z

217

Office of Nuclear Safety | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Safety Office of Nuclear Safety Organization Office of Health and Safety Office of Environmental Protection, Sustainability Support & Corporate Safety Analysis Office of...

218

Preliminary Safety Analysis Report for the Transuranic Storage Area Retrieval Enclosure at the Idaho National Engineering Laboratory. Revision 8  

SciTech Connect

This Transuranic Storage Area Retrieval Enclosure Preliminary Safety Analysis Report was completed as required by DOE Order 5480.23. The purpose of this document is to construct a safety basis that supports the design and permits construction of the facility. The facility has been designed to the requirements of a Radioactive Solid Waste Facility presented in DOE Order 6430.1A.

1993-03-01T23:59:59.000Z

219

Safety analysis of the CSTR-1 bench-scale coal liquefaction unit  

SciTech Connect

The objective of the program reported herein was to provide a Safety Analysis of the CSTR-1 bench scale unit located in Building 167 at the Pittsburgh Energy Technology Center. It was apparent that considerable effort was expended in the design and construction of the unit, and in the development of operating procedures, with regard to safety. Exhaust ventilation, H/sub 2/ and H/sub 2/S monitoring, overpressure protection, overtemperature protection, and interlock systems have been provided. Present settings on the pressure and temperature safety systems are too high, however, to insure prevention of vessel deformation or damage in all cases. While the occurrence of catastrophic rupture of a system pressure vessel (e.g., reactor, high pressure separators) is unlikely, the potential consequences to personnel are severe. Feasibility of providing shielding for these components should be considered. A more probable mode of vessel failure in the event of overpressure or overtemperature and failure of the safety system is yielding of the closure bolts followed by high pressure flow across the mating surfaces. As a minimum, shielding should be designed to restrict travel of resultant spray. The requirements for personal protective equipment are presently stated in rather broad and general terms in the operating procedures. Safe practices and procedures would be more assured if specific requirements were stated and included for each operational step. Recommendations were developed for all hazards triggered by the guidelines.

Hulburt, D.A.

1981-05-01T23:59:59.000Z

220

Analysis of batteries for use in photovoltaic systems. Final report  

SciTech Connect

An evaluation of 11 types of secondary batteries for energy storage in photovoltaic electric power systems is given. The evaluation was based on six specific application scenarios which were selected to represent the diverse requirements of various photovoltaic systems. Electrical load characteristics and solar insulation data were first obtained for each application scenario. A computer-based simulation program, SOLSIM, was then developed to determine optimal sizes for battery, solar array, and power conditioning systems. Projected service lives and battery costs were used to estimate life-cycle costs for each candidate battery type. The evaluation considered battery life-cycle cost, safety and health effects associated with battery operation, and reliability/maintainability. The 11 battery types were: lead-acid, nickel-zinc, nickel-iron, nickel-hydrogen, lithium-iron sulfide, calcium-iron sulfide, sodium-sulfur, zinc-chlorine, zinc-bromine, Redox, and zinc-ferricyanide. The six application scenarios were: (1) a single-family house in Denver, Colorado (photovoltaic system connected to the utility line); (2) a remote village in equatorial Africa (stand-alone power system); (3) a dairy farm in Howard County, Maryland (onsite generator for backup power); (4) a 50,000 square foot office building in Washington, DC (onsite generator backup); (5) a community in central Arizona with a population of 10,000 (battery to be used for dedicated energy storage for a utility grid-connected photovoltaic power plant); and (6) a military field telephone office with a constant 300 W load (trailer-mounted auxiliary generator backup). Recommendations for a research and development program on battery energy storage for photovoltaic applications are given, and a discussion of electrical interfacing problems for utility line-connected photovoltaic power systems is included. (WHK)

Podder, A.; Kapner, M.

1981-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Photovoltaic venture analysis. Final report. Volume II. Appendices  

DOE Green Energy (OSTI)

A description of the integrating model for photovoltaic venture analysis is given; input assumptions for the model are described; and the integrating model program listing is given. The integrating model is an explicit representation of the interactions between photovoltaic markets and supply under alternative sets of assumptions. It provides a consistent way of assembling and integrating the various assumptions, data, and information that have been obtained on photovoltaic systems supply and demand factors. Secondly, it provides a mechanism for understanding the implications of all the interacting assumptions. By representing the assumptions in a common, explicit framework, much more complex interactions can be considered than are possible intuitively. The integrating model therefore provides a way of examining the relative importance of different assumptions, parameters, and inputs through sensitivity analysis. Also, detailed results of model sensitivity analysis and detailed market and systems information are presented. (WHK)

Costello, D.; Posner, D.; Schiffel, D.; Doane, J.; Bishop, C.

1978-07-01T23:59:59.000Z

222

Photovoltaic venture analysis. Final report. Volume I. Executive summary  

DOE Green Energy (OSTI)

The objective of the study, government programs under investigation, and a brief review of the approach are presented. Potential markets for photovoltaic systems relevant to the study are described. The response of the photovoltaic supply industry is then considered. A model which integrates the supply and demand characteristics of photovoltaics over time was developed. This model also calculates the economic benefits associated with various government subsidy programs. Results are derived under alternative possible supply, demand, and macroeconomic conditions. A probabilistic analysis of the costs and benefits of a $380 million federal photovoltaic procurement initiative, as well as certain alternative strategies, is summarized. Conclusions and recommendations based on the analysis are presented.

Costello, D.; Posner, D.; Schiffel, D.; Doane, J.; Bishop, C.

1978-07-01T23:59:59.000Z

223

K Basin Hazard Analysis  

Science Conference Proceedings (OSTI)

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

PECH, S.H.

2000-08-23T23:59:59.000Z

224

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

SciTech Connect

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

Bruce Wright

2011-08-30T23:59:59.000Z

225

Market-analysis system for conservation technologies. Draft final report  

SciTech Connect

A prototype market analysis methodology to provide DOE decision makers guidance in evaluating and selecting strategies that promote energy conservation technologies is discussed. The methodology, named MASCOT (Market Analysis System for COnservation Technologies), was designed for the residential water heating market. However, the basic logic can be extended to other market segments, such as space heating and conditioning, and the commercial sector. MASCOT forecasts the market performance of any arbitrary set of technologies that the user chooses. The methodology captures the time-varying effects of technological and economic changes in the market, determines the critical features for new water heating technologies, calculates the likely energy impacts from the use of the actual technologies, and provides information concerning the sensitivity of the results to assumptions about market conditions, technology characteristics, and the factors underlying market penetration. (PSB)

Morris, P.A.; Thapa, M.N.; Bauman, D.S.; Froker, D.B.

1981-12-14T23:59:59.000Z

226

Economic impact analysis for the petroleum refineries NESHAP. Final report  

Science Conference Proceedings (OSTI)

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

NONE

1995-08-01T23:59:59.000Z

227

Advanced techniques for safety analysis applied to the gas turbine control system of ICARO co-generative plant  

E-Print Network (OSTI)

The paper describes two complementary and integrable approaches, a probabilistic one and a deterministic one, based on classic and advanced modelling techniques for safety analysis of complex computer based systems. The probabilistic approach is based on classical and innovative probabilistic analysis methods. The deterministic approach is based on formal verification methods. Such approaches are applied to the gas turbine control system of ICARO co generative plant, in operation at ENEA CR Casaccia. The main difference between the two approaches, behind the underlining different theories, is that the probabilistic one addresses the control system by itself, as the set of sensors, processing units and actuators, while the deterministic one also includes the behaviour of the equipment under control which interacts with the control system. The final aim of the research, documented in this paper, is to explore an innovative method which put the probabilistic and deterministic approaches in a strong relation to overcome the drawbacks of their isolated, selective and fragmented use which can lead to inconsistencies in the evaluation results. 1.

Ro Bologna; Ester Ciancamerla; Piero Incalcaterra; Michele Minichino; Andrea Bobbio; Universit Del Piemonte Orientale; Enrico Tronci

2001-01-01T23:59:59.000Z

228

New challenges in the safety analysis of DOE`s high-level waste tanks  

Science Conference Proceedings (OSTI)

Tank 241-SY-101, located at the Department of Energy Hanford Site, has periodically released up to 283 m{sup 3} (10,000 ft{sup 3}) of flammable gas. This release has been one of the highest priority DOE operational safety problems because of potential consequences if the gas were ignited during one of these releases. The gases include hydrogen and ammonia (fuels) and nitrous oxide (oxidizer). There have been many opinions regarding the controlling mechanisms for these releases, but demonstrating an adequate understanding of the problem, selecting a mitigation methodology, and preparing the safety analysis have presented numerous new challenges. The purpose of this report is to present an overview of the problem, the main issues, the method selected to mitigate this hazard, and the results of the mitigation program.

Edwards, J.N.; Pasamehmetoglu, K.O.; White, J.R. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

1994-09-01T23:59:59.000Z

229

2012 Annual Workforce Analysis and Staffing Plan Report - NNSA for Safety and Health  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1 1 Annual Workforce Analysis and Staffing Plan Report as of December 31, 2012 Reporting Office: NNSA NA-SH Section 1: Current Mission(s) of the Organization and Potential Changes The Office of the Associate Administrator for Safety and Health (NA-SH) provides mission enabling support to the NNSA Administrator, Central Technical Authority (CTA), Acquisition Executives, senior NNSA officials, program officers and site offices. NA-SH enables other NNSA organizations to fulfill NNSA missions while protecting the environment and safeguarding the safety and health of the public and the workforce. Section 2: SITE CHARACTERISTICS TABLE 1 Number of Hazard Category 1, 2, or 3 Nuclear Facilities: HC 1:_0_; HC 2: _0_; HC 3: _0_. Number of Radiological Facilities

230

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

Science Conference Proceedings (OSTI)

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

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

1994-02-01T23:59:59.000Z

231

Caucasus Seismic Information Network: Data and Analysis Final Report  

Science Conference Proceedings (OSTI)

The geology and tectonics of the Caucasus region (Armenia, Azerbaijan, and Georgia) are highly variable. Consequently, generating a structural model and characterizing seismic wave propagation in the region require data from local seismic networks. As of eight years ago, there was only one broadband digital station operating in the region an IRIS station at Garni, Armenia and few analog stations. The Caucasus Seismic Information Network (CauSIN) project is part of a nulti-national effort to build a knowledge base of seismicity and tectonics in the region. During this project, three major tasks were completed: 1) collection of seismic data, both in event catalogus and phase arrival time picks; 2) development of a 3-D P-wave velocity model of the region obtained through crustal tomography; 3) advances in geological and tectonic models of the region. The first two tasks are interrelated. A large suite of historical and recent seismic data were collected for the Caucasus. These data were mainly analog prior to 2000, and more recently, in Georgia and Azerbaijan, the data are digital. Based on the most reliable data from regional networks, a crustal model was developed using 3-D tomographic inversion. The results of the inversion are presented, and the supporting seismic data are reported. The third task was carried out on several fronts. Geologically, the goal of obtaining an integrated geological map of the Caucasus on a scale of 1:500,000 was initiated. The map for Georgia has been completed. This map serves as a guide for the final incorporation of the data from Armenia and Azerbaijan. Description of the geological units across borders has been worked out and formation boundaries across borders have been agreed upon. Currently, Armenia and Azerbaijan are working with scientists in Georgia to complete this task. The successful integration of the geologic data also required addressing and mapping active faults throughout the greater Caucasus. Each of the major faults in the region were identified and the probability of motion were assessed. Using field data and seismicity, the relative activity on each of these faults was determined. Furthermore, the sense of motion along the faults was refined using GPS, fault plane solutions, and detailed field studies. During the course of the integration of the active fault data, the existence of the proposed strike slip Borjomi-Kazbeki fault was brought into question. Although it had been incorporated in many active tectonic models over the past decade, field geologists and geophysicists in Georgia questioned its existence. Detailed field studies were carried out to determine the existence of the fault and estimate the slip along it; and it was found that the fault zone did not exist. Therefore, the convergence rate in the greater Caucasus must be reinterpreted in terms of thrust mechanisms, instead of strike-slip on the Borjomi-Kazbeki fault zone.

Randolph Martin; Mary Krasovec; Spring Romer; Timothy O'Connor; Emanuel G. Bombolakis; Youshun Sun; Nafi Toksoz

2007-02-22T23:59:59.000Z

232

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

Science Conference Proceedings (OSTI)

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

Not Available

1994-07-01T23:59:59.000Z

233

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

SciTech Connect

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

Not Available

1994-07-01T23:59:59.000Z

234

Analysis of leaded and unleaded gasoline pricing. Final report  

SciTech Connect

This report summarizes the evaluation of the cost price relation between the two fuels. The original scope of work identified three separate categories of effort: Gather and organize available data on the wholesale and retail prices of gasoline at a national level for the past 5 years. Using the data collected in Subtask 1, develop models of pricing practices that aid in explaining retail markups and price differentials for different types and grades of gasoline at different retail outlets in the current gasoline market. Using the data from Subtask 1 and the analysis framework from Subtask 2, analyze the likely range of future retail markups and price differentials for different grades of leaded and unleaded gasoline. The report is organized in a format that is different than suggested by the subtasks outlined above. The first section provides a characterization of the problem - data available to quantify cost and price of the fuels as well as issues that directly affect this relationship. The second section provides a discussion of issues likely to affect this relation in the future. The third section postulates a model that can be used to quantify the relation between fuels, octane levels, costs and prices.

1985-03-15T23:59:59.000Z

235

Analysis of photographic records of coal pyrolysis. Final report  

SciTech Connect

Bituminous coals upon heating undergo melting and pyrolytic decomposition with significant parts of the coal forming an unstable liquid that can escape from the coal by evaporation. The transient liquid within the pyrolyzing coal causes softening or plastic behavior that can influence the chemistry and physics of the process. Bubbles of volatiles can swell the softened coal mass in turn affecting the combustion behavior of the coal particles. The swelling behavior of individual coal particles has to be taken into account both as the layout as well as for the operation of pyrolysis, coking and performance of coal-fired boilers. Increased heating rates generally increase the amount of swelling although it is also known that in some cases, even highly swelling coals can be transformed into char with no swelling if they are heated slowly enough. The swelling characteristics of individual coal particles have been investigated by a number of workers employing various heating systems ranging from drop tube and shock tube furnaces, flow rate reactors and electrical heating coils. Different methods have also been employed to determine the swelling factors. The following sections summarize some of the published literature on the subject and outline the direction in which the method of analysis will be further extended in the study of the swelling characteristics of hvA bituminous coal particles that have been pyrolyzed with a laser beam.

Dodoo, J.N.D.

1991-10-01T23:59:59.000Z

236

Production cost analysis of Euphorbia lathyris. Final report  

DOE Green Energy (OSTI)

The purpose of this study is to estimate costs of production for Euphorbia lathyris (hereafter referred to as Euphorbia) in commercial-scale quantities. Selection of five US locations for analysis was based on assumed climatic and cultivation requirements. The five areas are: nonirrigated areas (Southeast Kansas and Central Oklahoma, Northeast Louisiana and Central Mississippi, Southern Illinois), and irrigated areas: (San Joaquin Valley and the Imperial Valley, California and Yuma, Arizona). Cost estimates are tailored to reflect each region's requirements and capabilities. Variable costs for inputs such as cultivation, planting, fertilization, pesticide application, and harvesting include material costs, equipment ownership, operating costs, and labor. Fixed costs include land, management, and transportation of the plant material to a conversion facility. Euphorbia crop production costs, on the average, range between $215 per acre in nonirrigated areas to $500 per acre in irrigated areas. Extraction costs for conversion of Euphorbia plant material to oil are estimated at $33.76 per barrel of oil, assuming a plant capacity of 3000 dry ST/D. Estimated Euphorbia crop production costs are competitive with those of corn. Alfalfa production costs per acre are less than those of Euphorbia in the Kansas/Oklahoma and Southern Illinois site, but greater in the irrigated regions. This disparity is accounted for largely by differences in productivity and irrigation requirements.

Mendel, D.A.

1979-08-01T23:59:59.000Z

237

Solar thermal repowering utility value analysis. Final report  

DOE Green Energy (OSTI)

The retrofit of solar central receiver energy supply systems to existing steam-electric generating stations (repowering) is being considered as a major programmatic thrust by DOE. The determination of a government response appropriate to the opportunities of repowering is an important policy question, and is the major reason for the analysis. The study objective is to define a government role in repowering that constitutes an efficient program investment in pursuit of viable private markets for heliostat-based energy systems. In support of that objective, the study is designed to identify the scope and nature of the repowering opportunity within the larger context of its contributions to central receiver technology development and commercialization. The Supply and Integration Tasks are documented elsewhere. This report documents the Demand Task, determining and quantifying the sources of the value of repowering and of central receiver technology in general to electric utilities. The modeling tools and assumptions used in the Demand Task are described and the results are presented and interpreted. (MCW)

Taylor, R.; Day, J.; Reed, B.; Malone, M.

1979-12-01T23:59:59.000Z

238

Production cost analysis of Euphorbia lathyris. Final report  

SciTech Connect

The purpose of this study is to estimate costs of production for Euphorbia lathyris (hereafter referred to as Euphorbia) in commercial-scale quantities. Selection of five US locations for analysis was based on assumed climatic and cultivation requirements. The five areas are: nonirrigated areas (Southeast Kansas and Central Oklahoma, Northeast Louisiana and Central Mississippi, Southern Illinois), and irrigated areas: (San Joaquin Valley and the Imperial Valley, California and Yuma, Arizona). Cost estimates are tailored to reflect each region's requirements and capabilities. Variable costs for inputs such as cultivation, planting, fertilization, pesticide application, and harvesting include material costs, equipment ownership, operating costs, and labor. Fixed costs include land, management, and transportation of the plant material to a conversion facility. Euphorbia crop production costs, on the average, range between $215 per acre in nonirrigated areas to $500 per acre in irrigated areas. Extraction costs for conversion of Euphorbia plant material to oil are estimated at $33.76 per barrel of oil, assuming a plant capacity of 3000 dry ST/D. Estimated Euphorbia crop production costs are competitive with those of corn. Alfalfa production costs per acre are less than those of Euphorbia in the Kansas/Oklahoma and Southern Illinois site, but greater in the irrigated regions. This disparity is accounted for largely by differences in productivity and irrigation requirements.

Mendel, D.A.

1979-08-01T23:59:59.000Z

239

Photosynthesis energy factory: analysis, synthesis, and demonstration. Final report  

DOE Green Energy (OSTI)

This quantitative assessment of the potential of a combined dry-land Energy Plantation, wood-fired power plant, and algae wastewater treatment system demonstrates the cost-effectiveness of recycling certain by-products and effluents from one subsystem to another. Designed to produce algae up to the limit of the amount of carbon in municipal wastewater, the algae pond provides a positive cash credit, resulting mainly from the wastewater treatment credit, which may be used to reduce the cost of the Photosynthesis Energy Factory (PEF)-generated electricity. The algae pond also produces fertilizer, which reduces the cost of the biomass produced on the Energy Plantation, and some gas. The cost of electricity was as low as 35 mills per kilowatt-hour for a typical municipally-owned PEF consisting of a 65-MWe power plant, a 144-acre algae pond, and a 33,000-acre Energy Plantation. Using only conventional or near-term technology, the most cost-effective algae pond for a PEF is the carbon-limited secondary treatment system. This system does not recycle CO/sub 2/ from the flue gas. Analysis of the Energy Plantation subsystem at 15 sites revealed that plantations of 24,000 to 36,000 acres produce biomass at the lowest cost per ton. The following sites are recommended for more detailed evaluation as potential demonstration sites: Pensacola, Florida; Jamestown, New York; Knoxville, Tennessee; Martinsville, Virginia, and Greenwood, South Carolina. A major possible extension of the PEF concept is to include the possibility for irrigation.

Not Available

1978-11-01T23:59:59.000Z

240

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

SciTech Connect

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.

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

2008-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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241

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

Science Conference Proceedings (OSTI)

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

Not Available

1994-08-01T23:59:59.000Z

242

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

SciTech Connect

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

Not Available

1994-08-01T23:59:59.000Z

243

Environment, Safety, and Health Assurance Program Standard: analysis in the context of Department of Energy environment, safety, and health requirements  

SciTech Connect

An Environment, Safety and Health (ES and H) Assurance Program Standard is presented. The Standard was published for comments in a previous document, and has been revised as a result of those comments. The benefits of using the Standard are described and, through the use of comparisons, it is demonstrated that the requirements of the applicable Department of Energy (DOE) Orders are satisfied by a program designed in accordance with the Standard.

Ellingson, A.C.; Trauth, C.A. Jr.

1979-09-01T23:59:59.000Z

244

Safety Bulletin  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Bulletin Bulletin Offtce 01 Health. Safety and Sa<:urtty Events Beyond Design Safety Basis Analysis No. 2011-01 PURPOSE This Safety Alert provides information on a safety concern related to the identification and mitigation of events that may fall outside those analyzed in the documented safety analysis. BACKGROUND On March 11 , 2011 , the Fukushima Daiichi nuclear power station in Japan was damaged by a magnitude 9.0 earthquake and the subsequent tsunami. While there is still a lot to be learned from the accident · about the adequacy of design specifications and the equipment failure modes, reports from the Nuclear Regulatory Commission (NRC) have identified some key aspects of the operational emergency at the Fukushima Daiichi nuclear power station.

245

Safety Bulletin  

NLE Websites -- All DOE Office Websites (Extended Search)

those analyzed in the documented safety analysis. BACKGROUND On March 11 , 2011 , the Fukushima Daiichi nuclear power station in Japan was damaged by a magnitude 9.0 earthquake and...

246

Understanding waste phenomenology to reduce the amount of sampling and analysis required to resolve Hanford waste tank safety issues  

SciTech Connect

Safety issues associated with Hanford Site waste tanks arose because of inadequate safety analyses and high levels of uncertainty over the release of radioactivity resulting from condensed phase exothermic chemical reactions (organic solvent fires, organic complexant-nitrate reactions, and ferrocyanide-nitrate reactions). The approach to resolving the Organic Complexant, Organic Solvent, and Ferrocyanide safety issues has changed considerably since 1990. The approach formerly utilized core sampling and extensive analysis of the samples with the expectation the data would provide insight into the hazard. This resulted in high costs and the generation of a large amount of data that was of limited value in resolving the safety issues. The new approach relies on an understanding of the hazard phenomenology to focus sampling and analysis on those analytes that are key to ensuring safe storage of the waste.

Meacham, J.E.; Babad, H.

1996-02-01T23:59:59.000Z

247

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

Science Conference Proceedings (OSTI)

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

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

1993-10-01T23:59:59.000Z

248

Guidance for the design and management of a maintenance plan to assure safety and improve the predictability of a DOE nuclear irradiation facility. Final report  

SciTech Connect

A program is recommended for planning the maintenance of DOE nuclear facilities that will help safety and enhance availability throughout a facility`s life cycle. While investigating the requirements for maintenance activities, a major difference was identified between the strategy suitable for a conventional power reactor and one for a research reactor facility: the latter should provide a high degree of predicted availability (referred to hereafter as ``predictability``) to its users, whereas the former should maximize total energy production. These differing operating goals necessitate different maintenance strategies. A strategy for scheduling research reactor facility operation and shutdown for maintenance must balance safety, reliability,and predicted availability. The approach developed here is based on three major elements: (1) a probabilistic risk analysis of the balance between assured reliability and predictability (presented in Appendix C), (2) an assessment of the safety and operational impact of maintenance activities applied to various components of the facility, and (3) a data base of historical and operational information on the performance and requirements for maintenance of various components. These factors are integrated into a set of guidelines for designing a new highly maintainable facility, for preparing flexible schedules for improved maintenance of existing facilities, and for anticipating the maintenance required to extend the life of an aging facility. Although tailored to research reactor facilities, the methodology has broader applicability and may therefore be used to improved the maintenance of power reactors, particularly in anticipation of peak load demands.

Booth, R.S.; Kryter, R.C.; Shepard, R.L.; Smith, O.L. [Oak Ridge National Lab., TN (United States); Upadhyaya, B.R. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Rowan, W.J.

1994-10-01T23:59:59.000Z

249

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

DOE Green Energy (OSTI)

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

Crowder, C.; Hurley, T.

1981-03-01T23:59:59.000Z

250

Report of study of tanker safety and pollution prevention requirements for U. S. tankers in domestic trade. Final report  

SciTech Connect

This report presents results of a study to determine if tanker safety and pollution prevention measures in addition to those contained in 1978 Protocols to SOLAS 74 and MARPOL 73 agreements should be applied to U.S. tank vessels in domestic trade. The study examined the risks associated with the marine transportation of oil by U.S. tank vessels in domestic trade, looking at the present and projected U.S. flag tank vessel fleet, oil movements by these vessels, and resulting hazards to people, property, and the marine environment. Possible preventative actions, including extension of ship construction and equipment requirements contained in 1978 Protocols to SOLAS 74 and MARPOL 73 to smaller tankships, were identified and examined. Estimates were made of: (1) the impact of possible preventative actions on accidental and operational oil discharges and damage to the marine environment, (2) tankship fires and explosions, and (3) transportation costs and capital requirements. On the basis of information presented in the study, a key recommendation is the adoption of additional measures to control oil discharges from possible future transportation of OCS oil to shore by U.S. tank vessels. Requirements for segregated ballast tanks or clean ballast tanks should not otherwise be extended to smaller U.S. tankships in domestic trade.

1978-06-01T23:59:59.000Z

251

Radiation Detection Scenario Analysis Toolbox (RADSAT) Test Case Implementation Final Report  

Science Conference Proceedings (OSTI)

Final report for the project. This project was designed to demonstrate the use of the Radiation Detection Scenario Analysis Toolbox (RADSAT) radiation detection transport modeling package (developed in a previous NA-22 project) for specific radiation detection scenarios important to proliferation detection.

Shaver, Mark W.

2010-09-27T23:59:59.000Z

252

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

SciTech Connect

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

NONE

1997-12-01T23:59:59.000Z

253

Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization  

SciTech Connect

The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. 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. We provide a users guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

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

2013-05-01T23:59:59.000Z

254

Guidance on health effects of toxic chemicals. Safety Analysis Report Update Program  

SciTech Connect

Martin Marietta Energy Systems, Inc. (MMES), and Martin Marietta Utility Services, Inc. (MMUS), are engaged in phased programs to update the safety documentation for the existing US Department of Energy (DOE)-owned facilities. The safety analysis of potential toxic hazards requires a methodology for evaluating human health effects of predicted toxic exposures. This report provides a consistent set of health effects and documents toxicity estimates corresponding to these health effects for some of the more important chemicals found within MMES and MMUS. The estimates are based on published toxicity information and apply to acute exposures for an ``average`` individual. The health effects (toxicological endpoints) used in this report are (1) the detection threshold; (2) the no-observed adverse effect level; (3) the onset of irritation/reversible effects; (4) the onset of irreversible effects; and (5) a lethal exposure, defined to be the 50% lethal level. An irreversible effect is defined as a significant effect on a person`s quality of life, e.g., serious injury. Predicted consequences are evaluated on the basis of concentration and exposure time.

Foust, C.B.; Griffin, G.D.; Munro, N.B.; Socolof, M.L.

1994-02-01T23:59:59.000Z

255

Safety analysis report vitrified high level waste type B shipping cask  

Science Conference Proceedings (OSTI)

This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

NONE

1995-03-01T23:59:59.000Z

256

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

Science Conference Proceedings (OSTI)

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

Dr. Bradley J Schrader

2009-03-01T23:59:59.000Z

257

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

SciTech Connect

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.

Dr. Bradley J Schrader

2010-10-01T23:59:59.000Z

258

Environment/Health/Safety (EHS): Laser Safety  

NLE Websites -- All DOE Office Websites (Extended Search)

Laser Safety Home Whom to Call Analysis of Laser Safety Occurrences: 2005-2011 Laser Bio-effects Laser Classification Laser Disposal Guide Laser Forms Laser Newsletter Laser Lab...

259

Safety Basis Information System  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis (SESA) SESA Home Mission & Functions Office of Sustainability, Environment, Safety and Anaylsis (SESA) Sustainability Support Environmental Policy & Assistance ...

260

September 26, 2011, Department letter transmitting the Implementation Plan for Board Recommendation 2010-1, Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers.  

NLE Websites -- All DOE Office Websites (Extended Search)

September 26, 2011 September 26, 2011 The Honorable Peter S. Winokur Chairman Defense Nuclear Facilities Safety Board 625 Indiana Avenue, NW, Suite 700 Washington, DC 20004-2941 Dear Mr. Chairman: Enclosed is the Department of Energy's Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 2010-1, Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers. This Plan provides the Department's approach for updating its Documented Safety Analysis Standards and requirements to clarify them in regards to performance of hazard and accident analysis and the identification of safety controls. I have assigned Dr. James B. O'Brien, Acting Director, Office of Nuclear Safety in the Office of Health, Safety and Security, as the Department's Responsible

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility  

SciTech Connect

The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.

Johnson, D.J.; Brehm, J.R.

1994-01-01T23:59:59.000Z

262

A path analysis of relationships among job stress, job satisfaction, motivation to transfer, and transfer of learning: perceptions of occupational safety and health administration outreach trainers  

E-Print Network (OSTI)

Many researchers have examined the effect of various work-related factors on transfer of learning. However, there has been little or no focus on the effect of key workplace factors such as job stress and job satisfaction on transfer of learning. The current study examines the relationship among job stress, job satisfaction, motivation to transfer and transfer of learning based on the perceptions of selected Occupational Safety and Health Administration (OSHA) outreach trainers who underwent training conducted by the Texas Engineering Extension, Texas. A 24-item questionnaire was utilized to collect data. The questionnaire was sent electronically to all outreach trainers who underwent the OSHA General Industry Course 501 during 2005, and the first six months of 2006. The sample included 418 respondents representing a population of 1234 outreach trainers. Descriptive statistics, Cronbachs alpha estimates for reliability, factor analysis, correlation analysis, regression analysis, path analysis, and Sobel tests were the analysis methods used in the study. The results from the analysis suggest that job stress and its related dimensions, time stress, and anxiety had an indirect correlation with transfer of learning through job satisfaction and motivation to transfer. Further, it was found that job stress, time stress, and anxiety predicted job satisfaction; time stress predicted anxiety; job satisfaction predicted motivation to transfer; and motivation to transfer predicted transfer of learning. Finally, path analysis results and mediation tests showed that: (1) the relationship between job stress and transfer was mediated by job satisfaction and motivation to transfer, (2) the relationship between time stress and transfer was mediated by job satisfaction and motivation to transfer, (3) the relationship between anxiety and transfer was mediated by job satisfaction and motivation to transfer, and finally (4) the relationship between time stress and transfer was mediated by anxiety, job satisfaction, and motivation to transfer.

Nair, Prakash Krishnan

2007-05-01T23:59:59.000Z

263

Letter from Nuclear Energy Institute regarding Integrated Safety Analysis: Why it is Appropropriate for Fuel Recycling Facilities  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

082 l F: 202.533.0166 l rxm@nei.org l www.nei.org 082 l F: 202.533.0166 l rxm@nei.org l www.nei.org Rod McCullum DIRECTOR FUEL CYCLE PROJECTS NUCLEAR GENERATION DIVISION September 10, 2010 Ms. Catherine Haney Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689 Dear Ms. Haney: 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 intended as an information source for the NRC and should serve as a foundation for discussion with industry representatives on the issue.

264

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

Science Conference Proceedings (OSTI)

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.

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.

2004-12-01T23:59:59.000Z

265

Challenges and methodology for safety analysis of a high-level waste tank with large periodic releases of flammable gas  

SciTech Connect

Tank 241-SY-101, located at the Department of Energy Hanford Site, has periodically released up to 10,000 ft{sup 3} of flammable gas. This release has been one of the highest-priority DOE operational safety problems. The gases include hydrogen and ammonia (fuels) and nitrous oxide (oxidizer). There have been many opinions regarding the controlling mechanisms for these releases, but demonstrating an adequate understanding of the problem, selecting a mitigation methodology, and preparing the safety analysis have presented numerous new challenges. The mitigation method selected for the tank was to install a pump that would mix the tank contents and eliminate the sludge layer believed to be responsible for the gas retention and periodic releases. This report will describe the principal analysis methodologies used to prepare the safety assessment for the installation and operation of the pump, and because this activity has been completed, it will describe the results of pump operation.

Edwards, J.N.; Pasamehmetoglu, K.O.; White, J.R. [Los Alamos National Lab., NM (United States); Stewart, C.W. [Pacific Northwest Lab., Richland, WA (United States)

1994-07-01T23:59:59.000Z

266

Radiation Safety  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Brotherhood of Locomotive Brotherhood of Locomotive Engineers & Trainmen Scott Palmer BLET Radiation Safety Officer New Hire Training New Hire study topics * GCOR * ABTH * SSI * Employee Safety * HazMat * Railroad terminology * OJT * 15-week class * Final test Hazardous Materials * Initial new-hire training * Required by OSHA * No specified class length * Open book test * Triennial module Locomotive Engineer Training A little bit older...a little bit wiser... * Typically 2-4 years' seniority * Pass-or-get-fired promotion * Intensive program * Perpetually tested to a higher standard * 20 Weeks of training * 15 of that is OJT * General Code of Operating Rules * Air Brake & Train Handling * System Special Instructions * Safety Instructions * Federal Regulations * Locomotive Simulators * Test Ride * Pass test with 90% Engineer Recertification

267

Nuclear Criticality Safety - Nuclear Engineering Division (Argonne...  

NLE Websites -- All DOE Office Websites (Extended Search)

Criticality Safety Nuclear Criticality Safety Overview Experience Analysis Tools Current NCS Activities Current R&D Activities DOE Criticality Safety Support Group (CSSG) Other...

268

An overview of severe accident modeling and analysis work for the ANS reactor conceptual safety analysis report  

Science Conference Proceedings (OSTI)

ORNL`s Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management has initiated severe accident analysis and related technology development efforts early-on in the design phase itself. Early consideration of severe accident issues will aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It will also provide a means for satisfying on- and off-site regulatory requirements and provide containment response and source term analyses for level-2 and -3 Probabilistic Risk Analyses (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions, and consequently provide insights for the development of strategies and design philosophies for accident management, mitigation, and emergency preparedness. This paper presents a perspective overview of the severe accident modeling and analysis work for the ANS Conceptual Safety Analysis Report (CSAR)

Taleyarkhan, R.P.

1992-12-31T23:59:59.000Z

269

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

SciTech Connect

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

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

2006-09-01T23:59:59.000Z

270

The Office of Health, Safety and Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Sustainability, Environment, Safety and Anaylsis (SESA) Sustainability Support Environmental Policy & Assistance Corporate Safety Programs Analysis Program Contacts...

271

Criticality Safety  

NLE Websites -- All DOE Office Websites (Extended Search)

Left Tab EVENTS Office of Nuclear Safety (HS-30) Office of Nuclear Safety Home Directives Nuclear and Facility Safety Policy Rules Nuclear Safety Workshops Technical...

272

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

SciTech Connect

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

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

2011-05-01T23:59:59.000Z

273

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

Science Conference Proceedings (OSTI)

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

Milliken, N.J., Westinghouse Hanford

1996-08-01T23:59:59.000Z

274

An analysis of the proposed MITR-III core to establish thermal-hydraulic limits at 10 MW. Final report  

Science Conference Proceedings (OSTI)

The 5 MW Massachusetts Institute of Technology Research Reactor (MITR-II) is expected to operate under a new license beginning in 1999. Among the options being considered is an upgrade in the heat removal system to allow operation at 10 MW. The purpose of this study is to predict the Limiting Safety System Settings and Safety Limits for the upgraded reactor (MITR-III). The MITR Multi-Channel Analysis Code was written to analyze the response of the MITR system to a series of anticipated transients in order to determine the Limiting Safety System Settings and Safety Limits under various operating conditions. The MIT Multi-Channel Analysis Code models the primary and secondary systems, with special emphasis placed on analyzing the thermal-hydraulic conditions in the core. The code models each MITR fuel element explicitly in order to predict the behavior of the system during flow instabilities. The results of the code are compared to experimental data from MITR-II and other sources. New definitions are suggested for the Limiting Safety System Settings and Safety Limits. MITR Limit Diagrams are included for three different heat removal system configurations. It is concluded that safe, year-round operating at 10 MW is possible, given that the primary and secondary flow rates are both increased by approximately 40%.

Harling, O.K.; Lanning, D.D.; Bernard, J.A.; Meyer, J.E.; Henry, A.F.

1997-06-01T23:59:59.000Z

275

FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter  

NLE Websites -- All DOE Office Websites (Extended Search)

H2 Safety Snapshot H2 Safety Snapshot Newsletter to someone by E-mail Share FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on Facebook Tweet about FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on Twitter Bookmark FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on Google Bookmark FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on Delicious Rank FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on Digg Find More places to share FCT Safety, Codes and Standards: H2 Safety Snapshot Newsletter on AddThis.com... Home Basics Current Approaches to Safety, Codes & Standards DOE Activities Quick Links Hydrogen Production Hydrogen Delivery Hydrogen Storage Fuel Cells Technology Validation Manufacturing Education Systems Analysis

276

A Comparison of the Safety Analysis Process and the Generation IV Proliferation Resistance/Physical Protection Assessment Methodology  

SciTech Connect

The Generation IV International Forum (GIF) is a vehicle for the cooperative international development of future nuclear energy systems. The Generation IV program has established primary objectives in the areas of sustainability, economics, safety and reliability, and Proliferation Resistance and Physical Protection (PR&PP). In order to help meet the latter objective a program was launched in December 2002 to develop a rigorous means to assess nuclear energy systems with respect to PR&PP. The study of Physical Protection of a facility is a relatively well established methodology, but an approach to evaluate the Proliferation Resistance of a nuclear fuel cycle is not. This paper will examine the Proliferation Resistance (PR) evaluation methodology being developed by the PR group, which is largely a new approach and compare it to generally accepted nuclear facility safety evaluation methodologies. Safety evaluation methods have been the subjects of decades of development and use. Further, safety design and analysis is fairly broadly understood, as well as being the subject of federally mandated procedures and requirements. It is therefore extremely instructive to compare and contrast the proposed new PR evaluation methodology process with that used in safety analysis. By so doing, instructive and useful conclusions can be derived from the comparison that will help to strengthen the PR methodological approach as it is developed further. From the comparison made in this paper it is evident that there are very strong parallels between the two processes. Most importantly, it is clear that the proliferation resistance aspects of nuclear energy systems are best considered beginning at the very outset of the design process. Only in this way can the designer identify and cost effectively incorporate intrinsic features that might be difficult to implement at some later stage. Also, just like safety, the process to implement proliferation resistance should be a dynamic, iterative process that continually evolves with the design.

T. A. Bjornard; M. D. Zentner

2006-05-01T23:59:59.000Z

277

A hazard analysis of human factors in safety-critical systems engineering  

Science Conference Proceedings (OSTI)

Safety incident studies often cite human factors as a major cause of accidents. At Bhopal in 1984 human error - the failure to follow safe operating procedures - instigated the deaths of thousands of people from cyanide poisoning. In this case, human ... Keywords: EN50128, IEC 61508, human factors, safety-critical systems engineering

Les Chambers

2006-04-01T23:59:59.000Z

278

Quality Assurance requirements - Safety Analysis Reports for Packaging. An effective approach in developing QA requirements  

SciTech Connect

Application of QA requirements for packaging and transportation of radioactive materials should not be solely based on safety-related considerations. The operability of items, components, and systems must be considered as equally important. The nuclear industry has begun to recognize operability considerations along with safety concerns. This has resulted in a new approach in establishing QA requirements for packaging.

Fabian, R.R.

1986-01-01T23:59:59.000Z

279

CY 2012 Annual Workforce Analysis and Staffing Plan - Chief of Nuclear Safety  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2 2 Reporting Office: Chief of Nuclear Safety . Section One: Current Mission(s) of the Organization and Potential Changes Revision 2 of U.S. Department of Energy Implementation Plan for DNFSB Recommendation 2004-1 established the seven core CTA responsibilities. The Office of the Chief of Nuclear Safety (CNS) performs to following functions in support of the CTA meeting these responsibilities: 1. Nuclear Safety Requirement Concurrence and Exemption * Concur with the determination of the applicability of DOE directives involving nuclear safety included in Energy and Science contracts pursuant to Department of Energy Acquisition Regulation (DEAR), 48 CFR 970.5204-2, Laws, regulations, and DOE directives, item (b). * Concur with nuclear safety requirements included in Energy and Science contracts pursuant to

280

CY 2011 Annual Workforce Analysis and Staffing Plan - Chief of Nuclear Safety  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1 1 Reporting Office: Chief of Nuclear Safety . Section One: Current Mission(s) of the Organization and Potential Changes Revision 2 of U.S. Department of Energy Implementation Plan for DNFSB Recommendation 2004-1 established the seven core CTA responsibilities. The Office of the Chief of Nuclear Safety (CNS) performs to following functions in support of the CTA meeting these responsibilities: 1. Nuclear Safety Requirement Concurrence and Exemption * Concur with the determination of the applicability of DOE directives involving nuclear safety included in Energy and Science contracts pursuant to Department of Energy Acquisition Regulation (DEAR), 48 CFR 970.5204-2, Laws, regulations, and DOE directives, item (b). * Concur with nuclear safety requirements included in Energy and Science contracts pursuant to

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module  

DOE Green Energy (OSTI)

The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

2006-07-01T23:59:59.000Z

282

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

SciTech Connect

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.

NONE

1994-10-01T23:59:59.000Z

283

Environment, Safety, and Health Risk Assessment Program (ESHRAP)  

SciTech Connect

The Environment, Safety and Health Risk Assessment Program (ESHRAP) models human safety and health risk resulting from waste management and environmental restoration activities. Human safety and health risks include those associated with storing, handling, processing, transporting, and disposing of radionuclides and chemicals. Exposures to these materials, resulting from both accidents and normal, incident-free operation, are modeled. In addition, standard industrial risks (falls, explosions, transportation accidents, etc.) are evaluated. Finally, human safety and health impacts from cleanup of accidental releases of radionuclides and chemicals to the environment are estimated. Unlike environmental impact statements and safety analysis reports, ESHRAP risk predictions are meant to be best estimate, rather than bounding or conservatively high. Typically, ESHRAP studies involve risk predictions covering the entire waste management or environmental restoration program, including such activities as initial storage, handling, processing, interim storage, transportation, and final disposal. ESHRAP can be used to support complex environmental decision-making processes and to track risk reduction as activities progress.

Eide, Steven Arvid; Thomas Wierman

2003-12-01T23:59:59.000Z

284

Analysis of a 1200 kV circuit breaker for a gas insulated substation. Final report  

Science Conference Proceedings (OSTI)

This report describes the work carried out to analyze and design a circuit for use in 1200 kV gas insulated substations. The first part of the project was devoted to a thorough analysis of the requirements for the circuit breaker from the standpoint of the electrical system in which it would operate. A conceptual design was selected and all of the components of the circuit breaker were designed, modeled and verified. Finally a plan was prepared for the construction of a complete circuit breaker.

Not Available

1985-01-01T23:59:59.000Z

285

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

SciTech Connect

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

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

2009-03-14T23:59:59.000Z

286

Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel  

SciTech Connect

So-called dedicated fuels will be utilized to obtain maximum transmutation and incineration rates of minor actinides (MAs) in accelerator-driven systems (ADSs). These fuels are characterized by a high-MA content and the lack of the classical fertile materials such as {sup 238}U or {sup 232}Th. Dedicated fuels still have to be developed; however, programs are under way for their fabrication, irradiation, and testing. In Europe, mainly the oxide route is investigated and developed. A dedicated core will contain multiple 'critical' fuel masses, resulting in a certain recriticality potential under core degradation conditions. The use of dedicated fuels may also lead to strong deterioration of the safety parameters of the reactor core, such as, e.g., the void worth, Doppler or the kinetics quantities, neutron generation time, and {beta}{sub eff}. Critical reactors with this kind of fuel might encounter safety problems, especially under severe accident conditions. For ADSs, it is assumed that because of the subcriticality of the system, the poor safety features of such fuels could be coped with. Analyses reveal some safety problems for ADSs with dedicated fuels. Additional inherent and passive safety measures are proposed to achieve the required safety level. A safety strategy along the lines of a defense approach is presented where these measures can be integrated. The ultimate goal of these measures is to eliminate any mechanistic severe accident scenario and the potential for energetics.

Maschek, Werner [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Rineiski, Andrei [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Flad, Michael [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Morita, Koji [Kyushu University Institute of Environmental Systems (Japan); Coste, Pierre [Commissariat a l'Energie Atomique CE Grenoble (France)

2003-02-15T23:59:59.000Z

287

Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor  

SciTech Connect

A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

Primm, Trent [ORNL; Gehin, Jess C [ORNL

2009-04-01T23:59:59.000Z

288

Vehicle Technologies Heavy Vehicle Program: FY 2007 Benefits Analysis, Methodology and Results - Final Report  

NLE Websites -- All DOE Office Websites (Extended Search)

7 Benefits Analysis, 7 Benefits Analysis, Methodology and Results - Final Report ANL-08/06 Energy Systems Division Availability of This Report This report is available, at no cost, at http://www.osti.gov/bridge. It is also available on paper to the U.S. Department of Energy and its contractors, for a processing fee, from: U.S. Department of Energy Office of Scientific and Technical Information P.O. Box 62 Oak Ridge, TN 37831-0062 phone (865) 576-8401 fax (865) 576-5728 reports@adonis.osti.gov Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States

289

Vehicle Technologies Heavy Vehicle Program: FY 2008 Benefit Analysis, Methodology and Results - Final Report  

NLE Websites -- All DOE Office Websites (Extended Search)

8 Benefits Analysis, 8 Benefits Analysis, Methodology and Results- Final Report ANL-08/07 Energy Systems Division Availability of This Report This report is available, at no cost, at http://www.osti.gov/bridge. It is also available on paper to the U.S. Department of Energy and its contractors, for a processing fee, from: U.S. Department of Energy Office of Scientific and Technical Information P.O. Box 62 Oak Ridge, TN 37831-0062 phone (865) 576-8401 fax (865) 576-5728 reports@adonis.osti.gov Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States

290

Microsoft Word - Final Report- Engineering-Economic Analysis of Syngas Storage.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

Engineering-Economic Analysis Engineering-Economic Analysis of Syngas Storage DOE/NETL-2008/1331 Final Report July 31, 2008 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference therein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States

291

Laser Safety Communiques  

NLE Websites -- All DOE Office Websites (Extended Search)

Argonne National Laboratory, July 17-19, 2007 Registration Form Workshop Agenda DOE Laser Safety Memo and Final Report, February 28, 2005 APS Laser OJT ANL CHM OJT Example...

292

Final Supplement Analysis for the Site-Wide Environmental Impact Statement for the Sandia National Laboratories, Sandia, New Mexico  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

81-SA-04 81-SA-04 Final Supplement Analysis for the Final Site-Wide Environmental Impact Statement for Sandia National Laboratories/New Mexico August 2006 U.S. Department of Energy National Nuclear Security Administration Sandia Site Office This page intentionally left blank COVER SHEET RESPONSIBLE AGENCY: U.S. DEPARTMENT OF ENERGY/NATIONAL NUCLEAR SECURITY ADMINISTRATION TITLE: Final Supplement Analysis for the Final Site-Wide Environmental Impact Statement for Sandia National Laboratories/New Mexico (DOE/EIS-0281-SA-04) CONTACT: For further information concerning this Supplement Analysis, contact Ms. Susan Lacy Environmental Team Leader Sandia Site Office National Nuclear Security Administration P. O. Box 5400 Albuquerque, New Mexico 87185-5400 Phone: (505) 845-5542

293

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

A Review of Light-Water Reactor Safety Studies," by A.V.due to a break in the reactor cooling cooling water the therecirculation - Failure of the reactor protection system.

Nero, A.V.

2010-01-01T23:59:59.000Z

294

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Addendum No. 1, through Revision b  

SciTech Connect

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 'Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision b', dated May 2007 (Addendum 1). The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The safety analysis of the package is documented in the 'Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package' (i.e., the Mound 1KW SARP, or the SARP). Addendum 1 incorporates a new fueled capsule assembly payload. The following changes have been made to add this payload: (1) The primary containment vessel (PCV) will be of the same design, but will increase in height to 11.16 in.; (2) A new graphite support block will be added to support up to three fueled capsule assemblies per package; (3) The cutting groove height on the secondary containment vessel (SCV) will be heightened to accommodate the taller PCV; and (4) A 3.38 in. high graphite filler block will be placed on top of the PCV. All other packaging features, as described in the Mound 1KW SARP [3], remain unchanged. This report documents the LLNL review of Addendum 1[1]. The specific review for each SARP Chapter is documented herein.

DiSabatino, A; West, M; Hafner, R; Russell, E

2007-10-04T23:59:59.000Z

295

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

SciTech Connect

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

Singh, M.; Energy Systems; TA Engineering

2008-02-29T23:59:59.000Z

296

NIF special equipment construction health and safety plan  

SciTech Connect

The purpose of this plan is to identify how the construction and deployment activities of the National Ignition Facility (NIF) Special Equipment (SE) will be safely executed. This plan includes an identification of (1) the safety-related responsibilities of the SE people and their interaction with other organizations involved; (2) safety related requirements, policies, and documentation; (3) a list of the potential hazards unique to SE systems and the mechanisms that will be implemented to control them to acceptable levels; (4) a summary of Environmental Safety and Health (ES&H) training requirements; and (5) requirements of contractor safety plans that will be developed and used by all SE contractors participating in site activities. This plan is a subsidiary document to the NIF Construction Safety Program (CSP) and is intended to compliment the requirements stated therein with additional details specific to the safety needs of the SE construction-related activities. If a conflict arises between these two documents, the CSP will supersede. It is important to note that this plan does not list all of the potential hazards and their controls because the design and safety analysis process is still ongoing. Additional safety issues win be addressed in the Final Safety Analysis Report, Operational Safety Procedures (OSPs), and other plans and procedures as described in Section 3.0 of this plan.

Sawicki, R.H.

1997-07-28T23:59:59.000Z

297

Review of Documented Safety Analysis Development for the Hanford Site Waste Treatment and Immobilization Plant (LBL Facilities), April 23, 2013 (HSS CRAD 45-58, Rev. 0)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. Department of U.S. Department of Energy Subject: Review of Documented Safety Analysis Development for the Hanford Site Waste Treatment and Immob ilization Plant (LBL Facilities) - C riteria and Review Approach D oc um~ HS: HSS CRAD 45-58 Rev: 0 Eff. Date: April 23, 2013 Office of Safety and Emergency Management Evaluations Acting Di rec or, Office of Safety and Emergency Nltanagement Evaluations Date: Apri l 23 , 20 13 Criteria and Review Approach Document ~~ trd,James Low Date: April 23 , 20 13 1.0 PURPOSE Within the Office of H.ealth, Safety and Security (HSS), the Office of Enforcement and Overs ight, Office of Safety and Emergency Management Evaluations (HS-45) miss io n is to assess the effectiveness of the environment, safety, health, and emergency management systems and practices used by line and

298

Nuclear Safety Regulatory Framework  

NLE Websites -- All DOE Office Websites (Extended Search)

Department of Energy 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 DOE Directives & Manuals DOE Standards Central Technical Authorities (CTA) Office of Health, Safety, and Security (HSS) Line Management SSO/ FAC Reps 48 CFR 970 48 CFR 952 Federal Acquisition Regulations External Oversight *Defense Nuclear Facility

299

Microsoft Word - Final EPIcode Guidance Report Version May 24 2004.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EH-4.2.1.3-EPIcode Code Guidance EH-4.2.1.3-EPIcode Code Guidance EPIcode Computer Code Application Guidance for Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 June 2004 EPIcode Guidance Report June 2004 Final Report ii INTENTIONALLY BLANK EPIcode Guidance Report June 2004 Final Report iii FOREWORD This document provides guidance to Department of Energy (DOE) facility analysts in the use of the EPIcode computer code for supporting Documented Safety Analysis applications. Information is provided herein that supplements information found in the EPIcode documentation provided by the code developer. EPIcode is one of six computer codes designated by the DOE Office of Environmental, Safety and Health as a toolbox code for safety

300

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MELCOR Computer Code Application Guidance for Leak Path Factor in Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health U.S. Department of Energy 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 MELCOR LPF Guidance May 2004 Final Report ii INTENTIONALLY BLANK MELCOR LPF Guidance May 2004 Final Report iii Foreword This document provides guidance to Department of Energy (DOE) facility analysts in the use of the MELCOR computer code for supporting Documented Safety Analysis applications. Information is provided herein that supplements information found in the MELCOR documentation provided by the code developer. MELCOR is one of six computer codes designated by DOE's Office of Environmental, Safety and Health as a toolbox code for safety

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301

Microsoft Word - Final MACCS2 Guidance Report June 30 2004.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MACCS2-Code Guidance MACCS2-Code Guidance MACCS2 Computer Code Application Guidance for Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 June 2004 MACCS2 Guidance Report June 2004 Final Report iii INTENTIONALLY BLANK MACCS2 Guidance Report June 2004 Final Report iv FOREWORD This document provides guidance to Department of Energy (DOE) facility analysts in the use of the MACCS2 computer code for supporting Documented Safety Analysis applications. Information is provided herein that supplements information found in the MACCS2 documentation provided by the code developer. MACCS2 is one of six computer codes designated by the DOE Office of Environmental, Safety and Health as a toolbox code for safety

302

Microsoft Word - Final ALOHA Guidance Report Version May 24 2004.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EH-4.2.1.3-ALOHA Code Guidance EH-4.2.1.3-ALOHA Code Guidance ALOHA Computer Code Application Guidance for Documented Safety Analysis Final Report U.S. Department of Energy Office of Environment, Safety and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 June 2004 ALOHA Guidance Report June 2004 Final Report ii INTENTIONALLY BLANK ALOHA Guidance Report June 2004 Final Report iii FOREWORD This document provides guidance to Department of Energy (DOE) facility analysts in the use of the ALOHA computer code for supporting Documented Safety Analysis applications. Information is provided herein that supplements information found in the ALOHA documentation provided by the code developer. ALOHA is one of six computer codes designated by DOE's Office of Environmental, Safety and Health as a toolbox code for safety

303

U. S. research safety vehicle (RSV) phase I program. Volume III. RSV characteristics and performance specifications. Final report, Jan 1974--Apr 1975  

SciTech Connect

Current passenger car usage patterns and factors influencing usage are analyzed and projections of usage patterns in the mid-1980's are made. Current available data on six categories of vehicle accidents are analyzed and projections made of national accident patterns in the mid-80's; the effect of potential reductions in these projections as a result of safety programs and other factors related to driving safety are estimated. Based on the usage and accident projections, the characteristics of an RSV (weighing under 3,000 lbs C.W.) for operation in the mid-1980 traffic environment are described. A recommended set of specifications for the RSV are developed considering the potential safety payoff accruing to an increased level of safety performance, the need for energy conservation, availability of material resources, and changes in vehicle mix. (An executive summary of this report is presented in Volume I).

Andon, J.; Dodson, E.; Khadilkar, A.; Olson, R.; Pauls, L.

1975-06-01T23:59:59.000Z

304

Safety Resources  

NLE Websites -- All DOE Office Websites (Extended Search)

Resources Print LBNLPub-3000: Health and Safety Manual Berkeley Lab safety guide, policies and procedures. Environment, Health, and Safety (EH&S) Staff Contact information for the...

305

Nuclear Safety  

Energy.gov (U.S. Department of Energy (DOE))

Nuclear Safety information site that provides assistance and resources to field elements in implementation of requirements and resolving nuclear safety, facility safety, and quality assurance issues.

306

Public Safety Network Requirements  

Science Conference Proceedings (OSTI)

... Usage scenario. ... imposed by public safety applications and usage scenarios is key in ... requirements as shown in Figure 2. This analysis was used as ...

2010-10-05T23:59:59.000Z

307

Safety Analysis Report for Packaging (SARP) for USA/5790/BLF (DOE-AL) and USA/5791/BLF (DOE-AL)  

Science Conference Proceedings (OSTI)

This revised Safety Analysis Report for Packaging (SARP) includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control of shipping containers. Much of the information was previously submitted to AEC/OSD/ALO and the Department of Transportation (DOT) and provided the basis for obtaining special permits DOT-SP-5790 and DOT-SP-5791 as well as the Interim Certificates of Compliance until the original SARP could be prepared and Certificates of Compliance issued by ERDA. This SARP revision incorporates information on certain design changes, the most significant of which relate to the inner container for the type 5790 package. Complete physical and technical descriptions of the packages are presented. Each package consists of a cylindrical steel inner container centered within an insulating steel drum assembly. The contents may be any radioactive materials which satisfy the requirements established in this SARP. A shipment of plutonium-238 in the form of a solid oxide is evaluated in this SARP as an example. The results of the nuclear criticality safety analysis show how much of the fissile isotopes may be shipped as Fissile Class I, II, or III for each container. Design and development considerations, the tests and evaluations required to prove the ability of the containers to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Mound Facility experience in using the containers, and copies of the DOE Certificates of Compliance are included. Internal reviews of the original and revised SARP's have been performed in compliance with the requirement of DOEM 5201-Part V.

Roome, L.G.; Watkins, R.A.; Bertram, R.E.; Kreider, H.B.

1980-01-25T23:59:59.000Z

308

Microsoft Word - DTRS56-05-T-0003-Benchmark Results Report- Final...  

NLE Websites -- All DOE Office Websites (Extended Search)

Pipeline & Hazardous Materials Safety Administration Pipeline Inspection Technologies Demonstration Report Pipeline Safety Research & Development Program Final 2 EXECUTIVE SUMMARY...

309

Analysis of battery storage for commercial buildings. Phase 1 final report  

SciTech Connect

The application of battery storage to load leveling by the utility user represents a new concept in energy management. TRW Energy Management Systems Division has studied the possibility of combining an energy management computer/control system with a lead-acid/power processor system and explored the feasibility of demonstrating power management at a government facility. Candidate sites in the Washington, D.C. metropolitan area were evaluated by analyzing demand curves for electricity. One site, the Department of the Treasury's Bureau of Printing and Engraving is recommended as the best of the sites evaluated. Analysis using estimated production system costs of $130/kW for power processors and $80/kWh for lead acid batteries indicates a payback of nine years. However, if the Department of Energy's cost goals for batteries and converters are achieved, a payback in less than four years is possible. Furthermore, coupling battery energy storage with conventional computer based energy management is projected to offer substantial reductions in utility bills. Payback from a production system in less than two years is predicted. System design is based on using present day technolgy where possible for the system components. Capacity for the system has been set at 1.1 MWh with a peak load capability of 600 kW. Preliminary specifications are supplied. Facility modification and system layout are presented, giving alternate placements for the batteries. Floor loading and system safety are two critical design parameters.

1978-09-01T23:59:59.000Z

310

Analysis of battery storage for commercial buildings. Phase 1 final report  

DOE Green Energy (OSTI)

The application of battery storage to load leveling by the utility user represents a new concept in energy management. TRW Energy Management Systems Division has studied the possibility of combining an energy management computer/control system with a lead-acid/power processor system and explored the feasibility of demonstrating power management at a government facility. Candidate sites in the Washington, D.C. metropolitan area were evaluated by analyzing demand curves for electricity. One site, the Department of the Treasury's Bureau of Printing and Engraving is recommended as the best of the sites evaluated. Analysis using estimated production system costs of $130/kW for power processors and $80/kWh for lead acid batteries indicates a payback of nine years. However, if the Department of Energy's cost goals for batteries and converters are achieved, a payback in less than four years is possible. Furthermore, coupling battery energy storage with conventional computer based energy management is projected to offer substantial reductions in utility bills. Payback from a production system in less than two years is predicted. System design is based on using present day technolgy where possible for the system components. Capacity for the system has been set at 1.1 MWh with a peak load capability of 600 kW. Preliminary specifications are supplied. Facility modification and system layout are presented, giving alternate placements for the batteries. Floor loading and system safety are two critical design parameters.

Not Available

1978-09-01T23:59:59.000Z

311

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

DOE Green Energy (OSTI)

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

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

2010-04-30T23:59:59.000Z

312

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

SciTech Connect

The U.S. Department of Energy (DOE) conducted a final quantitative analysis to assess whether buildings constructed according to the requirements of the American National Standards Institute (ANSI)/American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE)/Illuminating Engineering Society of North America (IESNA) Standard 90.1-2010 (ASHRAE Standard 90.1-2010, Standard 90.1-2010, or 2010 edition) would result in energy savings compared with buildings constructed to ANSI/ASHRAE/IESNA Standard 90.1-2007(ASHRAE Standard 90.1-2007, Standard 90.1-2007, or 2007 edition). The final analysis considered each of the 109 addenda to ASHRAE Standard 90.1-2007 that were included in ASHRAE Standard 90.1-2010. All 109 addenda processed by ASHRAE in the creation of Standard 90.1-2010 from Standard 90.1-2007 were reviewed by DOE, and their combined impact on a suite of 16 building prototype models in 15 ASHRAE climate zones was considered. Most addenda were deemed to have little quantifiable impact on building efficiency for the purpose of DOE's final determination. However, out of the 109 addenda, 34 were preliminarily determined to have a measureable and quantifiable impact. A suite of 240 computer energy simulations for building prototypes complying with ASHRAE 90.1-2007 was developed. These prototypes were then modified in accordance with these 34 addenda to create a second suite of corresponding building simulations reflecting the same buildings compliant with Standard 90.1-2010. The building simulations were conducted using the DOE EnergyPlus building simulation software. The resulting energy use from the complete suite of 480 simulation runs was then converted to energy use intensity (EUI, or energy use per unit floor area) metrics (Site EUI, Primary EUI, and energy cost intensity [ECI]) results for each simulation. For each edition of the standard, these EUIs were then aggregated to a national basis for each prototype using weighting factors based on construction floor area developed for each of the 15 U.S. climate zones using commercial construction data. When compared, the resulting weighted EUIs indicated that each of the 16 building prototypes used less energy under Standard 90.1-2010 than under Standard 90.1-2007 on a national basis when considering site energy, primary energy, or energy cost. The EUIs were also aggregated across building types to a national commercial building basis using the same weighting data. On a national basis, the final quantitative analysis estimated a floor-space-weighted national average reduction in new building energy consumption of 18.2 percent for source energy and 18.5 percent when considering site energy. An 18.2 percent savings in energy cost, based on national average commercial energy costs for electricity and natural gas, was also estimated.

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

2011-10-31T23:59:59.000Z

313

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

Science Conference Proceedings (OSTI)

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'.

Hudecek, V. [Technical University of Ostrava, Ostrava (Czech Republic)

2008-09-15T23:59:59.000Z

314

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

Science Conference Proceedings (OSTI)

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 cos

Enercon Services, Inc.

2011-03-14T23:59:59.000Z

315

Plutonium Finishing Plant safety evaluation report  

SciTech Connect

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.

Not Available

1995-01-01T23:59:59.000Z

316

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

SciTech Connect

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

B. C. Culp

2007-05-01T23:59:59.000Z

317

Microsoft Word - Final Rule  

NLE Websites -- All DOE Office Websites (Extended Search)

10 CFR, part 835 Docket No. HS-RM-09-835 RIN 1992-AA-45 Occupational Radiation Protection AGENCY: Office of Health, Safety and Security Department of Energy ACTION: Final Rule...

318

Preclosure radiological safety analysis for the exploratory shaft facilities; Yucca Mountain Site Characterization Project  

SciTech Connect

This study assesses which structures, systems, and components of the exploratory shaft facility (ESF) are important to safety when the ESF is converted to become part of the operating waste repository. The assessment follows the methodology required by DOE Procedure AP-6.10Q. Failures of the converted ESF during the preclosure period have been evaluated, along with other underground accidents, to determine the potential offsite radiation doses and associated probabilities. The assessment indicates that failures of the ESF will not result in radiation doses greater than 0.5 rem at the nearest unrestricted area boundary. Furthermore, credible accidents in other underground facilities will not result in radiation doses larger than 0.5 rem, even if any structure, system, or component of the converted ESF fails at the same time. Therefore, no structure, system, or component of the converted ESF is important to safety.

Ma, C.W.; Miller, D.D.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

1992-06-01T23:59:59.000Z

319

Enforcement Guidance Supplement 99-03, Limitation of 10 CFR 830 to Equipment Referenced in the Safety Analysis Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

for Enforcement Department of Energy Washington DC 20585 October 20, 1999 MEMORANDUM FOR: DOE and Contractor PAAA Coordinators FROM: R. Keith Christopher Director Office of Enforcement and Investigation SUBJECT: Enforcement Guidance Supplement 99-03: Limitation of 10 CFR Part 830 to Equipment Referenced in the Safety Analysis Report Recently this Office received a reply to a Preliminary Notice of Violation (PNOV), although not denying any facts or conclusions in the PNOV and agreeing to pay the full imposed Civil Penalty, included arguments that some of the equipment cited in the

320

Desired Characteristics for Next Generation Integrated Nuclear Safety Analysis Methods and Software  

Science Conference Proceedings (OSTI)

As a result of economic, environmental, and policy imperatives, it is envisioned that operation of the current fleet of commercial nuclear power plants NPPs will extend significantly beyond their original licensing periods. This objective can be achieved only if these plants continue to operate in a safe and cost-effective manner. The capability to perform detailed technical safety analyses of operational events either actual or postulated and desired operational enhancements such as power uprates will c...

2010-12-23T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

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

DOE Green Energy (OSTI)

The purpose of this Study is to provide an economic review of current and proposed changes to the Select Area Fishery Evaluation Project (SAFE or Project). The Study results are the information requested in comments made on the Project by a joint review dated March 2005 by the Northwest Power and Conservation Council (NPCC) Independent Scientific Review Panel (ISRP) and Independent Economic Analysis Board (IEAB). North et al. (2006) addressed technical questions about operations and plans, and this report contains the response information for comments concerning Project economics. This report can be considered an economic feasibility review meeting guidelines for cost-effective analysis developed by the IEAB (2003). It also contains other economic measurement descriptions to illustrate the economic effects of SAFE. The SAFE is an expansion of a hatchery project (locally called the Clatsop Economic Development Council Fisheries Project or CEDC) started in 1977 that released an early run coho (COH) stock into the Youngs River. The Youngs River entrance to the Columbia River at River Mile 12 is called Youngs Bay, which is located near Astoria, Oregon. The purpose of the hatchery project was to provide increased fishing opportunities for the in-river commercial fishing gillnet fleet. Instead of just releasing fish at the hatchery, a small scale net pen acclimation project in Youngs Bay was tried in 1987. Hirose et al. (1998) found that 1991-1992 COH broodstock over-wintered at the net pens had double the smolt-to-adult return rate (SAR) of traditional hatchery release, less than one percent stray rates, and 99 percent fishery harvests. It was surmised that smolts from other Columbia River hatcheries could be hauled to the net pens for acclimation and release to take advantage of the SAR's and fishing rates. Proposals were tendered to Bonneville Power Administration (BPA) and other agencies to fund the expansion for using other hatcheries smolts and other off-channel release sites. The BPA, who had been providing funds to the Project since 1982, greatly increased their financial participation for the experimental expansion of the net pen operations in 1993. Instead of just being a funding partner in CEDC operations, the BPA became a major financing source for other hatchery production operations. The BPA has viewed the 10 plus years of funding since then as an explorative project with two phases: a 'research' phase ending in 1993, and a 'development' phase ending in 2006. The next phase is referred to in proposals to BPA for continued funding as an 'establishment' phase to be started in 2007. There are three components of SAFE: (1) The CEDC owns and operates the net pens in the Columbia River estuary on the Oregon side. The CEDC also owns and operates a hatchery on the South Fork Klaskanine River. (2) There are many other hatcheries contributing smolts to the net pen operations. The present suite of hatcheries are operated by the Washington Department of Fish and Wildlife (WDFW) and Oregon Department of Fish and Wildlife (ODFW). The WDFW owns and operates the net pens at Deep River on the Washington side of the Columbia River. (3) The monitoring and evaluation (M&E) responsibilities are performed by employees of WDFW and ODFW. BPA provides funding for all three components as part of NPCC Project No. 199306000. The CEDC and other contributing hatcheries have other sources of funds that also support the SAFE. BPA's minor share (less than 10 percent) of CEDC funding in 1982 grew to about 55 percent in 1993 with the beginning of the development phase of the Project. The balance of the CEDC budget over the years has been from other federal, state, and local government programs. It has also included a 10 percent fee assessment (five percent of ex-vessel value received by harvesters plus five percent of purchase value made by processors) on harvests that take place in off-channel locations near the release sites. The CEDC total annual budget in the last several years has been in the $600 to $700 thousand range. The Project over

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

2006-11-01T23:59:59.000Z

322

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

Science Conference Proceedings (OSTI)

The purpose of this Study is to provide an economic review of current and proposed changes to the Select Area Fishery Evaluation Project (SAFE or Project). The Study results are the information requested in comments made on the Project by a joint review dated March 2005 by the Northwest Power and Conservation Council (NPCC) Independent Scientific Review Panel (ISRP) and Independent Economic Analysis Board (IEAB). North et al. (2006) addressed technical questions about operations and plans, and this report contains the response information for comments concerning Project economics. This report can be considered an economic feasibility review meeting guidelines for cost-effective analysis developed by the IEAB (2003). It also contains other economic measurement descriptions to illustrate the economic effects of SAFE. The SAFE is an expansion of a hatchery project (locally called the Clatsop Economic Development Council Fisheries Project or CEDC) started in 1977 that released an early run coho (COH) stock into the Youngs River. The Youngs River entrance to the Columbia River at River Mile 12 is called Youngs Bay, which is located near Astoria, Oregon. The purpose of the hatchery project was to provide increased fishing opportunities for the in-river commercial fishing gillnet fleet. Instead of just releasing fish at the hatchery, a small scale net pen acclimation project in Youngs Bay was tried in 1987. Hirose et al. (1998) found that 1991-1992 COH broodstock over-wintered at the net pens had double the smolt-to-adult return rate (SAR) of traditional hatchery release, less than one percent stray rates, and 99 percent fishery harvests. It was surmised that smolts from other Columbia River hatcheries could be hauled to the net pens for acclimation and release to take advantage of the SAR's and fishing rates. Proposals were tendered to Bonneville Power Administration (BPA) and other agencies to fund the expansion for using other hatcheries smolts and other off-channel release sites. The BPA, who had been providing funds to the Project since 1982, greatly increased their financial participation for the experimental expansion of the net pen operations in 1993. Instead of just being a funding partner in CEDC operations, the BPA became a major financing source for other hatchery production operations. The BPA has viewed the 10 plus years of funding since then as an explorative project with two phases: a 'research' phase ending in 1993, and a 'development' phase ending in 2006. The next phase is referred to in proposals to BPA for continued funding as an 'establishment' phase to be started in 2007. There are three components of SAFE: (1) The CEDC owns and operates the net pens in the Columbia River estuary on the Oregon side. The CEDC also owns and operates a hatchery on the South Fork Klaskanine River. (2) There are many other hatcheries contributing smolts to the net pen operations. The present suite of hatcheries are operated by the Washington Department of Fish and Wildlife (WDFW) and Oregon Department of Fish and Wildlife (ODFW). The WDFW owns and operates the net pens at Deep River on the Washington side of the Columbia River. (3) The monitoring and evaluation (M&E) responsibilities are performed by employees of WDFW and ODFW. BPA provides funding for all three components as part of NPCC Project No. 199306000. The CEDC and other contributing hatcheries have other sources of funds that also support the SAFE. BPA's minor share (less than 10 percent) of CEDC funding in 1982 grew to about 55 percent in 1993 with the beginning of the development phase of the Project. The balance of the CEDC budget over the years has been from other federal, state, and local government programs. It has also included a 10 percent fee assessment (five percent of ex-vessel value received by harvesters plus five percent of purchase value made by processors) on harvests that take place in off-channel locations near the release sites. The CEDC total annual budget in the last several years has been in the $600 to $700 thousand range. The Project over

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

2006-11-01T23:59:59.000Z

323

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network (OSTI)

L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

Galvez, Cristhian

2011-01-01T23:59:59.000Z

324

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

Science Conference Proceedings (OSTI)

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.

Not Available

1993-01-01T23:59:59.000Z

325

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

DOE Green Energy (OSTI)

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.

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

1992-07-01T23:59:59.000Z

326

Plasma analysis and diagnostics for high efficiency amorphous solar cell production. Final report  

DOE Green Energy (OSTI)

This is a project that sought to improve the amorphous silicon-germanium (SiGe) thin film deposition process in the production of solar cells. To accomplish this, the electron cyclotron resonance (ECR) plasma discharge, employed for the thin film deposition, was modified. Changes in the parameters of the plasma were monitored with diagnostic techniques, similar to those used in fusion plasma studies. That was the primary contribution from ORNL. Only one phase was contained in the statement of work, with the following tasks: (1) Develop a detailed program for plasma characterization. (2) Carry-out plasma modeling and analysis to support deposition systems design. (3) Operate experimental deposition systems for the purpose of plasma characterization. (4) Analyze data. (5) Modify deposition as directed by measurements. (6) This final report, which was deemed to be the only deliverable of this small project. And while the modified ECR discharge did not show measurable improvement of the conditions relevant to the deposition process, much was learned about the plasma parameters in the process. Some ideas on alternative designs are being discuss and funding options for testing such designed are being sought.

Klepper, C.C.

1994-12-21T23:59:59.000Z

327

Stair Safety  

NLE Websites -- All DOE Office Websites (Extended Search)

Stair Safety: Causes and Prevention of Stair Safety: Causes and Prevention of Residential Stair Injuries Cornell Department of Design & Cornell University Cooperative Environmental Analysis Martha Van Rensselaer Hall Extension 607-255-2144 Ithaca, NY 14853 In the United States during 1997 about 27,000 people were killed by unintentional home injuries. 1 Figure 1 illustrates the causes of some of the injuries that resulted in death. As you can see, falls account for the majority of incidents. Also in 1997, 6.8 million people suffered home accidents that resulted in disabling injuries. 1 While data on the number of injuries related to stairs and steps are not available for 1997, data from 1996 show that 984,000 people experienced injuries related to home stairs or steps during

328

Integrated Yucca Mountain Safety Case and Supporting Analysis: EPRI's Phase 7 Performance Assessment  

Science Conference Proceedings (OSTI)

After approval of the Yucca Mountain Site Recommendation by the President and Congress in 2001, the U.S. Department of Energy (DOE) entered the construction pre-license application phase with the U.S. Nuclear Regulatory Commission (NRC). A successful license application for the proposed spent fuel and high level waste repository at Yucca Mountain depends on a robust demonstration of long-term safety. It also depends on prioritizing the work left to do in a stepwise manner consistent with the particular p...

2002-12-29T23:59:59.000Z

329

Nuclear criticality safety analysis summary report: The S-area defense waste processing facility  

SciTech Connect

The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality.

Ha, B.C.

1994-10-21T23:59:59.000Z

330

Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Rel ated Applications  

Science Conference Proceedings (OSTI)

This report provides methodology that can be used to perform safety classification of non-process computer programs, such as design and analysis tools, that are not resident or embedded (installed as part of) plant systems, structures, and components. The report also provides guidance for using commercial-grade dedication methodology to accept commercially procured computer programs that perform a safety-related function. The guidance is intended for use by subject matter experts in the acceptance of com...

2012-06-04T23:59:59.000Z

331

Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications: Revision 1 of 1025243  

Science Conference Proceedings (OSTI)

This report supersedes EPRI 1025243 and provides methodology that can be used to perform safety classification of non-process computer programs, such as design and analysis tools, that are not resident or embedded (installed as part of) plant systems, structures, and components. The report also provides guidance for using commercial-grade dedication methodology to accept commercially procured computer programs that perform a safety-related function. The guidance is intended for use by subject matter ...

2013-12-19T23:59:59.000Z

332

An OSHA based approach to safety analysis for nonradiological hazardous materials  

SciTech Connect

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.

Yurconic, M.

1992-08-01T23:59:59.000Z

333

An OSHA based approach to safety analysis for nonradiological hazardous materials  

SciTech Connect

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.

Yurconic, M.

1992-08-01T23:59:59.000Z

334

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

Science Conference Proceedings (OSTI)

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.

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

1999-12-01T23:59:59.000Z

335

Environment/Health/Safety (EHS)  

NLE Websites -- All DOE Office Websites (Extended Search)

S S A B C D E F G H I J K L M N O P Q R S T U V W X Y Z SAAR - Supervisor's Accident Analysis Report SAAR for Division Safety Coordinators Safety Concerns/Comments Safety Engineering (Division) Safety Committee Safety Advisory Committee (LBNL) Safety Coordinator and Liaison Resources Safety Flicks Safety Shoes Safety Walk Around Check List Safety Walk Around Check List for Managers Satellite Accumulation Areas Security call x5472 Security and Emergency Operations Shipping & Transporting Hazardous Materials Shoemobile (schedule) (form) Site Access (parking permits, gate passes, buses) Site Environmental Report Site Map SJHA Spot Award Program Stop Work Policy Stretch Break Software-RSIGuard Subcontractor Job Hazard Analysis

336

Safety System Oversight  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety System Oversight Safety System Oversight Office of Nuclear Safety Home Safety System Oversight Home Annual SSO/FR Workshop DOE Safety Links › ORPS Info › Operating Experience Summary › DOE Lessons Learned › Accident Investigation Program Assessment Tools › SSO CRADS Subject Matter Links General Program Information › Program Mission Statement › SSO Program Description › SSO Annual Award Program › SSO Annual Award › SSO Steering Committee › SSO Program Assessment CRAD SSO Logo Items Site Leads and Steering Committee Archive Facility Representative Contact Us HSS Logo SSO SSO Program News Congratulations to Ronnie L. Alderson of Nevada Field Office, the Winner of the 2012 Safety System Oversight Annual Award! 2012 Safety System Oversight Annual Award Nominees SSO Staffing Analysis

337

Chemical Safety Program  

NLE Websites -- All DOE Office Websites (Extended Search)

Program Program Home Chemical Safety Topical Committee Library Program Contacts Related Links Site Map Tools 2013 Chemical Safety Workshop Archived Workshops Contact Us Health and Safety HSS Logo Chemical Safety Program logo The Department of Energy's (DOE's) Chemical Safety web pages provide a forum for the exchange of best practices, lessons learned, and guidance in the area of chemical management. This page is supported by the Chemical Safety Topical Committee which was formed to identify chemical safety-related issues of concern to the DOE and pursue solutions to issues identified. Noteworthy products are the Chemical Management Handbooks and the Chemical Lifecycle Cost Analysis Tool, found under the TOOLS menu. Chemical Management Handbook Vol (1) Chemical Management Handbook Vol (2)

338

Safety and Technical Services  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety and Technical Services Safety and Technical Services Minimize The Safety and Technical Services (STS) organization is a component of the Office of Science's (SC's) Oak Ridge Integrated Support Center. The mission of STS is to provide excellent environmental, safety, health, quality, and engineering support to SC laboratories and other U.S. Department of Energy program offices. STS maintains a full range of technically qualified Subject Matter Experts, all of whom are associated with the Technical Qualifications Program. Examples of the services that we provide include: Integrated Safety Management Quality Assurance Planning and Metrics Document Review Tracking and trending analysis and reporting Assessments, Reviews, Surveillances and Inspections Safety Basis Support SharePoint/Dashboard Development for Safety Programs

339

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

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Analysis of Coconut-Derived Analysis of Coconut-Derived Biodiesel and Conventional Diesel Fuel Samples from the Philippines Task 2 Final Report T.L. Alleman and R.L. McCormick Milestone Report NREL/MP-540-38643 January 2006 National Renewable Energy Laboratory 1617 Cole Boulevard, Golden, Colorado 80401-3393 303-275-3000 * www.nrel.gov Operated for the U.S. Department of Energy Office of Energy Efficiency and Renewable Energy by Midwest Research Institute * Battelle Contract No. DE-AC36-99-GO10337 Analysis of Coconut- Derived Biodiesel and Conventional Diesel Fuel Samples from the Philippines Task 2 Final Report T.L. Alleman and R.L. McCormick Prepared under Task Nos. WF3Y.1000 and FC02.0800 under an agreement between the U.S. Agency for International Development

340

A Duration Analysis of Food Safety Recall Events in the United States: January, 2000 to October, 2009  

E-Print Network (OSTI)

The safety of the food supply in the United States has become an issue of prominence in the minds of ordinary Americans. Several government agencies, including the United States Department of Agriculture and the Food and Drug Administration, are charged with the responsibility of preserving the safety of the food supply. Food is withdrawn from the market in a product recall when tainted or mislabeled and has the potential to harm the consumer in some manner. This research examines recall events issued by firms over the period of January, 2000 through October, 2009 in the United States. Utilizing economic and management theory to establish predictions, this study employs the Cox proportional hazard regression model to analyze the effects of firm size and branding on the risk of recall recurrence. The size of the firm was measured in both billions of dollars of sales and in thousands of employees. Branding by the firm was measured as a binary variable that expressed if a firm had a brand and as a count of the number of brands within a firm. This study also provides a descriptive statistical analysis and several findings based on the recall data specifically relating to annual occurrences, geographical locations of the firms involved, types of products recalled, and reasons for recall. We hypothesized that the increasing firm size would be associated with increased relative risk of a recall event while branding and an increasing portfolio of brands would be associated with decreased relative risk of a recall event. However, it was found that increased firm size and branding by the firm are associated with an increased risk of recall occurrence. The results of this research can have implications on food safety standards in both the public and private sectors.

Joy, Nathaniel Allen

2010-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Enforcement Regulations and Directives - Worker Safety and Health...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Worker Safety and Health Enforcement Regulations and Directives - Worker Safety and Health Regulations 10 C.F.R. Part 851 - Worker Safety and Health Program; Final Rule 10 C.F.R....

342

Safety-Basis Thermal Analysis for KE Basin Sludge Transport and Storage  

DOE Green Energy (OSTI)

A series of safety-basis thermal and gas generation analyses were completed and independently reviewed to assess the thermal performance of a large diameter container (LDC) containing KE Basin sludge. The results demonstrate: (1) the sludge transport system (STS) containing a LDC can safely transport a KE basin sludge payload up to 2.0 m{sup 3} and, (2) large diameter containers with sludge payloads up to 2.0 m{sup 3} can be safely stored in a process cell at T Plant. The transport and storage analyses are based on a conservative set of assumptions, including limiting environmental conditions. Conclusions drawn from the transport and storage results were not impacted by changes in the radial gap between the cask and LDC, purge gas (i.e., either helium or nitrogen), sludge porosity, or thermal conductivity. The design of the transport cask and large diameter container can accommodate reasonable changes in these values. Both transport from KE Basin and long-term storage at T Plant are addressed for sludge payloads up to 2.0 m{sup 3}. Additional analyses determined the expected range of T Plant environmental temperatures, the hydrogen and oxygen generation rate due to the radiolysis of water, and the maximum hydrogen concentration within a process cell due to chemical reactions and the radiolysis of water. All sludge temperature and hydrogen concentration criteria for transport and storage are met. The analyses assumed a safety-basis sludge mixture defined as 60% by volume floor and 40% by volume canister sludge with 35% retained gas, and a conservative segregated (axial) distribution of metallic uranium (resulting from particulate settling) with associated safety-basis properties. The analyses recognized that the retrieval process would produce non-uniform sludge distributions. Four batch process loadings of 0.5m{sup 3} each are assumed. Each process batch loading will settle and segregate (separate) into two layers: an active layer containing all the metallic uranium which is chemically active, and a non-active layer containing uranium oxide, non-uranium material, and no metallic uranium. This is a conservative representation of operational controls designed to limit the metallic uranium concentration. The sludge layers are assumed to remain intact during transport and storage.

HEARD, F.J.; SATHYANARAYANA, J.J.

2002-09-30T23:59:59.000Z

343

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

Science Conference Proceedings (OSTI)

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

NONE

1995-04-01T23:59:59.000Z

344

Pipeline Safety  

Science Conference Proceedings (OSTI)

Pipeline Safety. Summary: Our goal is to provide standard test methods and critical data to the pipeline industry to improve safety and reliability. ...

2012-11-13T23:59:59.000Z

345

Preparation Guide for U. S. Department of Energy Nonreator Nuclear Facility Document Safety Analysis  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SENSITIVE DOE-STD-3009-94 July 1994 CHANGE NOTICE NO. 1 January 2000 CHANGE NOTICE NO. 2 April 2002 DOE STANDARD PREPARATION GUIDE FOR U.S DEPARTMENT OF ENERGY NONREACTOR NUCLEAR FACILITY DOCUMENTED SAFETY ANALYSES U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS TS This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161;

346

Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analysis  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE-STD-3009-94 July 1994 CHANGE NOTICE NO. 12 January 2000 5 December 24 April 20021 DOE STANDARD PREPARATION GUIDE FOR U.S DEPARTMENT OF ENERGY NONREACTOR NUCLEAR FACILITY DOCUMENTED SAFETY ANALYSISANALYSES REPORTS U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161;

347

Flammable gas tank safety program: Technical basis for gas analysis and monitoring  

DOE Green Energy (OSTI)

Several Hanford waste tanks have been observed to exhibit periodic releases of significant quantities of flammable gases. Because potential safety issues have been identified with this type of waste behavior, applicable tanks were equipped with instrumentation offering the capability to continuously monitor gases released from them. This document was written to cover three primary areas: (1) describe the current technical basis for requiring flammable gas monitoring, (2) update the technical basis to include knowledge gained from monitoring the tanks over the last three years, (3) provide the criteria for removal of Standard Hydrogen Monitoring System(s) (SHMS) from a waste tank or termination of other flammable gas monitoring activities in the Hanford Tank farms.

Estey, S.D.

1998-04-22T23:59:59.000Z

348

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

DOE Green Energy (OSTI)

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.

Crandall, R.S.; Nelson, B.P. (National Renewable Energy Lab., Golden, CO (United States)); Moskowitz, P.D.; Fthenakis, V.M. (Brookhaven National Lab., Upton, NY (United States))

1992-07-01T23:59:59.000Z

349

Thermal reactor safety  

SciTech Connect

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.

1980-06-01T23:59:59.000Z

350

Three dimensional effects in analysis of PWR steam line break accident  

E-Print Network (OSTI)

A steam line break accident is one of the possible severe abnormal transients in a pressurized water reactor. It is required to present an analysis of a steam line break accident in the Final Safety Analysis Report (FSAR) ...

Tsai, Chon-Kwo

351

Initial Northwest Power Act Power Sales Contracts : Final Environmental Impact Statement. Volume 1, Environmental Analysis.  

DOE Green Energy (OSTI)

This is volume 1 of the final environmental impact statement of the Bonneville Power Administration Information is included on the following: Purpose of and need for action; alternatives including the proposed action; affected environment; and environmental consequences.

United States. Bonneville Power Administration.

1992-01-01T23:59:59.000Z

352

Implementation of Revision 19 of the TRUPACT-II Safety Analysis Report at Rocky Flats Environmental Technology Site  

Science Conference Proceedings (OSTI)

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.

D'Amico, E.; O'Leary, J.; Bell, S.; Djordjevic, S.; Givens, C,; Shokes, T.; Thompson, S.; Stahl, S.

2003-02-25T23:59:59.000Z

353

DOE-STD-1027-92; Hazard Categorization and Accident Analysis Techniques For Compliance With DOE Order 5480.23, Nuclear Safety Analysis Reports  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

7-92 7-92 December 1992 CHANGE NOTICE NO.1 September 1997 DOE STANDARD HAZARD CATEGORIZATION AND ACCIDENT ANALYSIS TECHNIQUES FOR COMPLIANCE WITH DOE ORDER 5480.23, NUCLEAR SAFETY ANALYSIS REPORTS U.S. Department of Energy AREA SAFT Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; (423) 576-8401. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161; (703) 487-4650. Order No. DE98001283 Change Notice No. 1 DOE-STD-1027-92

354

From Crowd Dynamics to Crowd Safety: A Video-Based Analysis  

E-Print Network (OSTI)

The study of crowd dynamics is interesting because of the various self-organization phenomena resulting from the interactions of many pedestrians, which may improve or obstruct their flow. Besides formation of lanes of uniform walking direction and oscillations at bottlenecks at moderate densities, it was recently discovered that stop-and-go waves [D. Helbing et al., Phys. Rev. Lett. 97, 168001 (2006)] and a phenomenon called "crowd turbulence" can occur at high pedestrian densities [D. Helbing et al., Phys. Rev. E 75, 046109 (2007)]. Although the behavior of pedestrian crowds under extreme conditions is decisive for the safety of crowds during the access to or egress from mass events as well as for situations of emergency evacuation, there is still a lack of empirical studies of extreme crowding. Therefore, this paper discusses how one may study high-density conditions based on suitable video data. This is illustrated at the example of pilgrim flows entering the previous Jamarat Bridge in Mina, 5 kilometers ...

Johansson, Anders; Al-Abideen, Habib Z; Al-Bosta, Salim

2008-01-01T23:59:59.000Z

355

Style, content and format guide for writing safety analysis documents: Volume 2, Safety assessment reports for DOE non-nuclear facilities  

SciTech Connect

The purpose of Volume 2 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Assessment Reports (SAs) for DOE non-nuclear facilities at Sandia National Laboratories. The scope of Volume 2 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SAs for DOE non-nuclear facilities.

Mahn, J.A.; Silver, R.C.; Balas, Y.; Gilmore, W.

1995-07-01T23:59:59.000Z

356

Wind energy mission analysis. Final report, appendices A--J. [USA  

DOE Green Energy (OSTI)

Information is presented concerning meteorological data and supporting analyses, gross energy consumption patterns and end-use analysis, analysis for industrial applications of wind energy conversion systems (WECS), analysis for residential applications of WECS, analysis for application of WECS to communities remote from utility grids, analysis for agricultural applications of WECS, regional evaluation of the economics of wind turbine generation to the U. S. electric utility district, impact of storage on WECS, financial analysis techniques, and system spacing.

Not Available

1977-02-18T23:59:59.000Z

357

Lift truck safety review  

SciTech Connect

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.

Cadwallader, L.C.

1997-03-01T23:59:59.000Z

358

Final Safety Evaluation Report to license the construction and operation of a facility to receive, store, and dispose of 11e.(2) byproduct material near Clive, Utah (Docket No. 40-8989)  

Science Conference Proceedings (OSTI)

The Final Safety Evaluation Report (FSER) summarizes the US Nuclear Regulatory Commission (NRC) staff`s review of Envirocare of Utah, Inc.`s (Envirocare`s) application for a license to receive, store, and dispose of uranium and thorium byproduct material (as defined in Section 11e.(2) of the Atomic Energy Act of 1954, as amended) at a site near Clive, Utah. Envirocare proposes to dispose of high-volume, low-activity Section 11e.(2) byproduct material in separate earthen disposal cells on a site where the applicant currently disposes of naturally occurring radioactive material (NORM), low-level waste, and mixed waste under license by the Utah Department of Environmental Quality. The NRC staff review of the December 23, 1991, license application, as revised by page changes dated July 2 and August 10, 1992, April 5, 7, and 10, 1993, and May 3, 6, 7, 11, and 21, 1993, has identified open issues in geotechnical engineering, water resources protection, radon attenuation, financial assurance, and radiological safety. The NRC will not issue a license for the proposed action until Envirocare adequately resolves these open issues.

Not Available

1994-01-01T23:59:59.000Z

359

Evolution of Safety Basis Documentation for the Fernald Site  

SciTech Connect

The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure.

Brown, T.; Kohler, S.; Fisk, P.; Krach, F.; Klein, B.

2004-03-01T23:59:59.000Z

360

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

Reports and Publications (EIA)

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

Robert T. Eynon

1996-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Technical Review Report for the Safety Analysis Report for Packaging Model 9977 S-SARP-G-00001 Revision 2  

SciTech Connect

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.

DiSabatino, A; Hafner, R; West, M

2007-10-04T23:59:59.000Z

362

Safety, Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety, Security Safety, Security Safety, Security LANL'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 work Our commitments We conduct our work safely and responsibly to achieve our mission. We ensure a safe and healthful environment for workers, contractors, visitors, and other on-site personnel. We protect the health, safety, and welfare of the general public. We do not compromise safety for personal, programmatic, or

363

Nuclear Safety Management  

NLE Websites -- All DOE Office Websites (Extended Search)

[6450-01-P] [6450-01-P] DEPARTMENT OF ENERGY 10 CFR Part 830 Nuclear Safety Management AGENCY: Department of Energy (DOE). ACTION: Final Rule. SUMMARY: The Department of Energy (DOE) is issuing a final rule regarding Nuclear Safety Management. This Part establishes requirements for the safe management of DOE contractor and subcontractor work at the Department's nuclear facilities. Today's rule adopts the sections that will make up the generally applicable provisions for Part 830. It also adopts the specific section on provisions for developing and implementing a formalized quality assurance program. EFFECTIVE DATE: This regulation becomes effective [insert 30 days after publication in the Federal Register.] FOR FURTHER INFORMATION CONTACT: Frank Hawkins, U.S. Department of Energy, Nuclear Safety

364

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

DOE Green Energy (OSTI)

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.

Crandall, R.S.; Nelson, B.P. [National Renewable Energy Lab., Golden, CO (United States); Moskowitz, P.D.; Fthenakis, V.M. [Brookhaven National Lab., Upton, NY (United States)

1992-07-01T23:59:59.000Z

365

Safety of high speed guided ground transportation systems: Comparison of magnetic and electric fields of conventional and advanced electrified transportation systems. Final report, September 1992-March 1993  

Science Conference Proceedings (OSTI)

Concerns exist regarding the potential safety, environmental and health effects on the public and on transportation workers due to electrification along new or existing rail corridors, and to proposed maglev and high speed rail operations. Therefore, the characterization of electric and magnetic fields (EMF) produced by both steady (dc) and alternating currents (ac) at power frequency (50 Hz in Europe and 60 Hz in the U.S.) and above, in the Extreme Low Frequency (ELF) range (3-3000 Hz) is of interest. The report summarizes and compares the results of a survey of EMF characteristics (spatial, temporal and frequency bands) for representative conventional railroad and transit and advanced high-speed systems including: the German TR-07 maglev system; the Amtrak Northeast Corridor (NEC) and North Jersey Transit (NJT) trains; the Washington, DC Metrorail (WMATA) and the Boston, MA (MBTA) transit systems; and the French TGV-A high speed rail system. This comprehensive comparative EMF survey produced both detailed data and statistical summaries of EMF profiles, and their variability in time and space. EMF ELF levels for WMATA are also compared to those produced by common environmental sources at home, work, and under power lines, but have specific frequency signatures.

Dietrich, F.M.; Feero, W.E.; Jacobs, W.L.

1993-08-01T23:59:59.000Z

366

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

STRUCTURE-SOIL- 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 displacements for EFB and PB * In-plane- shear from SASSI at EFB wall 2 Building Description 3 The Process Building is a massive reinforced concrete structure supported approximately 40 feet below the finished grade. The PB approximate foundation dimensions are approximately

367

Fusion Engineering and Design 80 (2006) 111137 ARIES-AT safety design and analysis  

E-Print Network (OSTI)

-AT coolant and structuresa Material Li17Pb83 SiC ORNL FS W 304 SS Inconel-625 Density (g/cm3) 9.58 3.22 7. / Fusion Engineering and Design 80 (2006) 111­137 115 Table 2 (Continued ) Material Li17Pb83 SiC ORNL FS W%, indicating that the sim- pler Approx 1 method could be used in future analysis without introducing large

California at San Diego, University of

368

ENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION  

E-Print Network (OSTI)

SERVICES ENVIRONMENTAL HEALTH & SAFETY Discovery 2 Building, Room 265 8888 University Drive BurnabyENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION SIMON FRASER UNIVERSITY SAFETY & RISK SIGNAGE 26740 INCIDENT INVESTIGATION Supervisors, Safety Committees, EHS LABORATORY SAFETY 27265

369

Technical Review Report for the Model 9975-96 Package Safety Analysis Report for Packaging (S-SARP-G-00003, Revision 0, January 2008)  

SciTech Connect

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.

West, M

2009-05-22T23:59:59.000Z

370

Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor  

SciTech Connect

The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

Wilson, G.E.

1992-01-01T23:59:59.000Z

371

Safety Bulletins  

NLE Websites -- All DOE Office Websites (Extended Search)

2009-01: Sulfur Hexafluoride Awareness Safety Bulletin 2008-03: Reporting Work-Related Heart Attacks Safety Bulletin 2008-02: Quality Assurance Concern at Wright Industries, Inc....

372

2012 Annual Workforce Analysis and Staffing Plan Report - NNSA for Safety and Health - NA-26  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1 Annual Workforce Analysis and Staffing Plan Report Draft as of December 31, 2012 Reporting Office: _NA-26 Office of Fissile Material Disposition at SRS____ Section 1: Current Mission(s) of the Organization and Potential Changes 1. The Office of Fissile Material Disposition (NA-26) is part of the National Nuclear Security Administration (NNSA). NA-26 supports NNSA Strategic Plan Goal #2, "Provide technical leadership to limit or prevent the spread of materials, technology, and expertise relating to weapons of mass destruction; advance the technologies to detect the proliferation of weapons of mass destruction worldwide, and eliminate or secure inventories of surplus materials and infrastructure usable for nuclear weapons." The NA-26 organization focuses on the safe and secure disposition of

373

Dependability analysis of a safety critical system the LHC beam dumping system at CERN  

E-Print Network (OSTI)

This thesis presents the dependability study of the Beam Dumping System of the Large Hadron Collider (LHC), the high energy particle accelerator to be commissioned at CERN in summer 2007. There are two identical, independent LHC Beam Dumping Systems (LBDS), one per LHC beam, each consisting of a series of magnets that extract the particle beam from the LHC ring into the extraction line leading to the absorbing block. The consequences of a failure within the LBDS can be very severe. This risk is reduced by applying redundancy to the design of the most critical components and on-line surveillance that, in case of a detected failure, issues a safe operation abort, called false beam dump. The system has been studied applying Failure Modes Effects and Criticality Analysis (FMECA) and reliability prediction. The system failure processes have been represented with a state transition diagram, governed by a Markov regenerative stochastic process, and analysed for different operational scenarios for one year of operati...

Filippini, R

2006-01-01T23:59:59.000Z

374

Near Term Hybrid Passenger Vehicle Development Program. Phase I, Final report. Appendix D: sensitivity analysis  

DOE Green Energy (OSTI)

This report on the Sensitivity of Mission Analysis and Trade-off Studies provides an analysis of the sensitivity of the results of previous mission analysis and performance specification studies to the possible variations of the values of significant parameters as projected to the year 1985. These parameters include vehicle usage by purpose, driving cycles, trip lengths, ownership projections, and life-cycle costs. Tabulated data are included from calculations with variations in these parameters. (LCL)

Traversi, M.

1979-07-03T23:59:59.000Z

375

Hazard analysis of compressed natural gas fueling systems and fueling procedures used at retail gasoline service stations. Final report  

Science Conference Proceedings (OSTI)

An evaluation of the hazards associated with operations of a typical compressed natural gas (CNG) fueling station is presented. The evaluation includes identification of a typical CNG fueling system; a comparison of the typical system with ANSI/NFPA (American National Standards Institute/National Fire Protection Association) Standard 52, Compressed Natural Gas (CNG) Vehicular Fuel System, requirements; a review of CNG industry safety experience as identified in current literature; hazard identification of potential internal (CNG system-specific causes) and external (interface of co-located causes) events leading to potential accidents; and an analysis of potential accident scenarios as determined from the hazard evaluation. The study considers CNG dispensing equipment and associated equipment, including the compressor station, storate vessels, and fill pressure sensing system.

NONE

1995-04-28T23:59:59.000Z

376

Analysis of a flow metering device for low-quality steam-water flows. Final report  

DOE Green Energy (OSTI)

The goal of this project is to investigate the potential of the meter configuration consisting of a sharp-edged contraction section followed by an extended length of constant area duct and finally a diffuser section for pressure recovery. This and two other configurations were tested. These configurations and the reasons underlying their selection are described and discussed. It is concluded that Murdock's correlation for steam/water flow through orifices and sudden contraction sections at low qualities is invalid and the metering scheme based on it is inoperative. (MHR)

Crowe, C.T.

1979-06-26T23:59:59.000Z

377

Geothermal evaluation and analysis of the Yucca Mountain Repository, Nevada; Final report, July 1, 1989--December 31, 1989  

DOE Green Energy (OSTI)

This is the final report on the geothermal analysis and evaluation for the proposed nuclear waste repository at Yucca Mountain, for the period of July 1, 1989 to December 31, 1989. Heat flow values were derived by measuring the thermal conductivities of samples taken from selected exploratory wells. Temperature gradients are recorded from the same wells. By using computer generated contour maps of the area, another interpretation of the heat-flow can be derived. Results of the mapping do not coincide with the past observations of the data. Another method used to evaluate the heat-flow of Yucca Mountain was to compare the temperature-depth relationship of the area. (MB)

NONE

1989-06-17T23:59:59.000Z

378

Mechanical Engineering Safety Note: Analysis and Control of Hazards Associated with NIF Capacitor Module Events  

Science Conference Proceedings (OSTI)

The NIF capacitor module was reviewed with respect to pressure venting and shrapnel containment during failures. A modified module concept was proposed that would adequately vent the pressure, yet be effective at containing shrapnel. Two large vents are provided on each side of the module. These have fixed vent areas, and are immediately accessible for pressure venting at the beginning of a pressure transient. A shrapnel shield is located on the outside of each vent opening forming a chute. The chute contains a collimator. This increases the number of bounces that shrapnel must take on the way out, and directs the shrapnel to the trap beneath. The trap contains a depth of clear pine, sufficient to completely absorb the energy of even the most energetic fragment considered. Based on a review of the evidence from past capacitor failures at the FANTM facility at Sandia National Laboratory, Albuquerque, and additional theoretical estimates, the peak pressure generated in the module during explosive events was estimated to be less than 40 psig. This internal pressure in the FANTM module appears to be tolerable, as only minor damage to the module and to internal components was observed after events. The new module concept proposed here provides increased venting area, fully available at the initiation of an event. It is expected that even less damage would be observed if an event occurred in a module with this design. The module joints and connections were formally reviewed with respect to their tolerance to a brief internal pressure as high as 40 psig. With minor modifications that have been incorporated into the design, the module was shown to maintain its integrity during such events. Some of the calculations performed estimated the quantity of dielectric oil that could be involved in a capacitor failure. It was determined that a very small amount of the available oil would contribute to the explosive event, on the order of 500 g or less. This is a small fraction of the total free oil available in a capacitor (approximately 10,900 g), on the order of 5% or less. The estimates of module pressure were used to estimate the potential overpressure in the capacitor bays after an event. It was shown that the expected capacitor bay overpressure would be less than the structural tolerance of the walls. Thus, it does not appear necessary to provide any pressure relief for the capacitor bays. The ray tracing analysis showed the new module concept to be 100% effective at containing fragments generated during the events. The analysis demonstrated that all fragments would impact an energy absorbing surface on the way out of the module. Thus, there is high confidence that energetic fragments will not escape the module. However, since the module was not tested, it was recommended that a form of secondary containment on the walls of the capacitor bays (e.g., 1.0 inch of fire-retardant plywood) be provided. Any doors to the exterior of the capacitor bays should be of equivalent thickness of steel or suitably armed with a thickness of plywood. Penetrations in the ceiling of the interior bays (leading to the mechanical equipment room) do not require additional protection to form a secondary barrier. The mezzanine and the air handling units (penetrations lead directly to the air handling units) provide a sufficient second layer of protection.

Brereton, S

2001-08-01T23:59:59.000Z

379

Safety Communications  

NLE Websites -- All DOE Office Websites (Extended Search)

Communications Communications New Staff & Guests Safety Topics ISM Plan Safety Communications Questions about safety and environmental compliance should first be directed to your supervisor or work lead. The Life Sciences Division Safety Coordinator Scott Taylor at setaylor@lbl.gov , 486-6133 (office), or (925) 899-4355 (cell); and Facilities Manager Peter Marietta at PMarietta@lbl.gov, 486-6031 (office), or 967-6596 (cell), are also sources of information. Your work group has a representative to the Division Environment, Health, & Safety Committee. This representative can provide safety guidance and offer a conduit for you to pass on your concerns or ideas. A list of current representatives is provided below. Additional safety information can be obtained on-line from the Berkeley Lab

380

Safety Analysis Report for Packaging (SARP): Model AL-M1 nuclear packaging (DOE C of C No. USA/9507/BLF)  

Science Conference Proceedings (OSTI)

This Safety Analysis Report for Packaging (SARP) satisfies the request of the US Department of Energy for a formal safety analysis of the shipping container identified as USA/9507/BLF, also called AL-M1, configuration 5. This report makes available to all potential users the technical information and the limits pertinent to the construction and use of the shipping containers. It includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. A complete physical and technical description of the package is presented. The package consists of an inner container centered within an insulated steel drum. The configuration-5 package contains tritiated water held on sorbent material. There are two other AL-M1 packages, designated configurations 1 and 3. These use the same insulated outer drum, but licensing of these containers will not be addressed in this SARP. Design and development considerations, the tests and evaluations required to prove the ability of the container to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Monsanto Research Corporation-Mound experience in using the containers, and a copy of the DOE/OSD/ALO Certificate of Compliance are included.

Coleman, H.L.; Whitney, M.A.; Williams, M.A.; Alexander, B.M.; Shapiro, A.

1987-11-24T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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381

Onshore permitting systems analysis for coal, oil, gas, geothermal and oil shale leases. Final report  

Science Conference Proceedings (OSTI)

The magnitude and complexity of permit processes raises a question as to their impact on the rate and scope of industrial development activity. One particular area where this issue is of concern is in new energy extraction and development activities. The initiation of new energy projects has been a national priority for several years. But, energy projects, because of their potential for creating land disturbances, are subject to many environmental and other regulations. Because of this, the permitting required of energy resource developers is extensive. Within the energy field, a major portion of development activities occurs on federal lands. This is particularly true in the Rocky Mountain states and Alaska where the principal landholder is the federal government. The permitting requirements for federal lands' development differ from those for private lands. This report assesses the impact of permitting processes for energy resource development on federal lands. The permitting processes covered include all of the major environmental, land-use, and safety permits required by agencies of federal and state governments. The lands covered include all federal lands, with emphasis on eight states with major development activities.

Not Available

1982-09-01T23:59:59.000Z

382

Formation and emission of methane in rice soils: Experimental determination and modeling analysis. Final report  

DOE Green Energy (OSTI)

Rice paddy soils have been identified as a major source of methane emissions contributing to the observed atmospheric increase in methane. This points to the need for a method of quantifying and predicting methane emissions for the widely varying conditions used in rice agriculture throughout the world. In the present work, a mathematical model for estimating the emission of methane from rice paddy soils is developed and refined. Kinetic parameters for methanogenesis in a Louisiana rice soil are determined from laboratory data on methane production from acetic acid substrate. Use of a stirred reactor allows simultaneous measurement of acetate consumption and methane production while minimizing mass transfer limitations. An existing model for rice plant growth is utilized to provide data on the availability of root exudates as a carbon source for the methanogens. The final methane model includes the kinetic parameters, plant data, and estimated transport parameters. With adjustments in these parameters, it provides an acceptable match to field data.

Law, V.J.; Bhattacharya, S.K.

1993-08-31T23:59:59.000Z

383

Consumer demand analysis: solar heating and cooling of buildings. Final report  

DOE Green Energy (OSTI)

This study concerns the acceptability of solar heating and cooling to homebuyers for residential applications. The study assesses the extent of homeowner awareness of solar technologies, estimates the acceptability of elevated first costs including willingness to trade higher initial costs for life cycle savings, and investigates the impact of solar aesthetics. Also explored are other areas of potential concern to homeowners in evaluating a solar alternative as well as positive motivations that would encourage purchase. Finally, the socioeconomic and attitudinal characteristics of individuals more likely to purchase a solar home rather than a conventional home were studied. The results are based on group depth interviews and personal interviews with active homeseekers, top executives of large residential development firms, and architects. The sample was split evenly between Denver, Colorado and the Philadelphia, Pa./Wilmington, Del. areas. Implications of the results for the commercialization of solar energy and possible public policy decisions are also discussed.

Scott, J.E.

1976-09-01T23:59:59.000Z

384

Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [ODSP-3 code; OTEC Steady-State Analysis Program  

DOE Green Energy (OSTI)

The following appendices are included; Dynamic Simulation Program (ODSP-3); sample results of dynamic simulation; trip report - NH/sub 3/ safety precautions/accident records; trip report - US Coast Guard Headquarters; OTEC power system development, preliminary design test program report; medium turbine generator inspection point program; net energy analysis; bus bar cost of electricity; OTEC technical specifications; and engineer drawings. (WHK)

Not Available

1978-12-04T23:59:59.000Z

385

Energy Engineering Analysis Program (EEAP), Laundry Plant Study, Fort Leonard Wood, Missouri. Volume II. Final report  

SciTech Connect

The purpose of this energy engineering analysis program (EEAP), Laundry Plant Study at Ft. Leonard Wood, Missouri is to develop energy saving type projects for funding through the energy conservation investment program (ECIP) or other applicable funding source. The following outlines the tasks performed in this study. The complete scope of work is included in Appendix G of this report. (1) Review of previously completed energy engineering analysis program (EEAP) studies applicable to the laundry facilities. (2) Perform a detailed survey of the laundry facility and associated energy using equipment. (3) Perform a complete energy audit and analysis of the laundry facilities. (4) Identify energy conservation opportunities including low cost/no cost items. (5) Provide complete programming and implementation documentation for all recommended ECO`s. (6) Prepare a comprehensive report documenting the work accomplished and the results of the study.

1992-02-28T23:59:59.000Z

386

California geothermal resource development environmental implications for ERCDC Environmental Analysis Office. Final report  

DOE Green Energy (OSTI)

The results of an analysis of the environmental implications for ERCDC Environmental Analysis Office (EAO) in relation to the development of California's geothermal resources are reported. While focusing primarily on environmental implications, particularly the natural, social, and economic elements, the report includes some ERCDC-wide policy and program considerations. The primary thrusts of the work have been in the development of an understanding of the interagency and intergovernmental environmental data and data-management roles and responsibilities and in the formulation of recommendations related thereto. Five appendices are included, one of which is a tax credit agreement between a power company and Skagit County, Washington. (JGB)

Roberts, J.A.

1977-02-01T23:59:59.000Z

387

Safety Advisories  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety Advisories Safety Advisories 2010 2010-08 Safety Advisory - Software Quality Assurance Firmware Defect in Programmable Logic Controller 2010-07 Safety Advisory - Revised Counterfeit Integrated Circuits Indictment 2010-06 Safety Advisory - Counterfeit Integrated Circuits Indictment 2010-05 Safety Advisory - Contact with Overhead Lines and Ground Step Potential 2010-04 Update - Leaking Acetylene Cylinder Shutoff Valves 2010-03 - Software Quality Assurance Microsoft Excel Software Issue 2010-02 - Leaking Acetylene Cylinder Shutoff Valves 2010-01 Update - Defective Frangible Ammunition 2009 2009-05 Software Quality Assurance - Errors in MACCS2 x/Q Calculations 2009-04 Update - SEELER Exothermic Torch 2009-03 - Defective Frangible Ammunition 2009-02 - Recall of Defense Technology Distraction Devices

388

Final Report for 'An Abstract Job Handling Grid Service for Dataset Analysis'  

Science Conference Proceedings (OSTI)

For Phase I of the Job Handling project, Tech-X has built a Grid service for processing analysis requests, as well as a Graphical User Interface (GUI) client that uses the service. The service is designed to generically support High-Energy Physics (HEP) experimental analysis tasks. It has an extensible, flexible, open architecture and language. The service uses the Solenoidal Tracker At RHIC (STAR) experiment as a working example. STAR is an experiment at the Relativistic Heavy Ion Collider (RHIC) at the Brookhaven National Laboratory (BNL). STAR and other experiments at BNL generate multiple Petabytes of HEP data. The raw data is captured as millions of input files stored in a distributed data catalog. Potentially using thousands of files as input, analysis requests are submitted to a processing environment containing thousands of nodes. The Grid service provides a standard interface to the processing farm. It enables researchers to run large-scale, massively parallel analysis tasks, regardless of the computational resources available in their location.

David A Alexander

2005-07-11T23:59:59.000Z

389

Analysis and comparison of biomass pyrolysis/gasification condensates: Final report  

DOE Green Energy (OSTI)

This report provides results of chemical and physical analysis of condensates from eleven biomass gasification and pyrolysis systems. The samples were representative of the various reactor configurations being researched within the Department of Energy, Biomass Thermochemical Conversion program. The condensates included tar phases and aqueous phases. The analyses included gross compositional analysis (elemental analysis, ash, moisture), physical characterization (pour point, viscosity, density, heat of combustion, distillation), specific chemical analysis (gas chromatography/mass spectrometry, infrared spectrophotometry, proton and carbon-13 nuclear magnetic resonance spectrometry) and biological activity (Ames assay and mouse skin tumorigenicity tests). These results are the first step of a longer term program to determine the properties, handling requirements, and utility of the condensates recovered from biomass gasification and pyrolysis. The analytical data demonstrates the wide range of chemical composition of the organics recovered in the condensates and suggests a direct relationship between operating temperature and chemical composition of the condensates. A continuous pathway of thermal degradation of the tar components as a function of temperature is proposed. Variations in the chemical composition of the organic components in the tars are reflected in the physical properties of tars and phase stability in relation to water in the condensate. The biological activity appears to be limited to the tars produced at high temperatures. 56 refs., 25 figs., 21 tabs.

Elliott, D.C.

1986-06-01T23:59:59.000Z

390

Operations research and systems analysis of geopressured/geothermal resources in Texas. Final report  

DOE Green Energy (OSTI)

A preliminary resource assessment, based on the best available parameters, was made to identify potentially suitable fairways. Of those examined only the Brazoria Fairway in the Frio Formation was able to produce sufficient fluid to meet the minimum requirements. These requirements are based upon the need for a well to produce an initial flow rate of 40,000 bbl/day with a 6% decline rate over a 30 year production period. Next, a development planning analysis was done to determine the number of wells that would have to be drilled in the fairway, considering the probability of success, and the number of drilling rigs available. The results of this analysis provided a time phased scenario and costs of developing the fairway. These were next used in an economic analysis. The economic analysis was performed to determine the present worth of using the resource under a range of values for the key economic parameters. The results of this study indicate that the commercial development of geopressured, geothermal resource is highly dependent upon the pricing of natural gas in the US, the development of tax incentives to spur development, and a better understanding of the nature of the resource through additional well tests.

Lesso, W.G.; Zinn, C.D.; Cornwell, J.

1981-05-01T23:59:59.000Z

391

October 24, 2003, Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

4.1 4.1 Revision 3 October 24, 2003 U. S. Department of Energy Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities October 24, 2003 CRAD - 4.2.4.1 Revision 3 October 24, 2003 ii TABLE OF CONTENTS ACRONYMS ..................................................................................................................................iii GLOSSARY ...................................................................................................................................iv 1.0 INTRODUCTION ...............................................................................................................1 2.0 BACKGROUND .................................................................................................................2

392

FreedomCAR and vehicle technologies heavy vehicle program FY 2006. Benefits analysis : methodology and results - final report.  

SciTech Connect

This report describes the approach to estimating benefits and the analysis results for the Heavy Vehicle Technologies activities of the Freedom Car and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identification of technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in the activities planned for FY 06. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. These benefits estimates, along with market penetrations and other results, are then modeled as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY06 Budget Request.

Singh, M.; Energy Systems; TA Engineering, Inc.

2006-01-31T23:59:59.000Z

393

Freedom car and vehicle technologies heavy vehicle program : FY 2007 benefits analysis, methodology and results -- final report.  

SciTech Connect

This report describes the approach to estimating the benefits and analysis results for the Heavy Vehicle Technologies activities of the FreedomCar and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identifying technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in subsequent activities. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. The market penetrations are used as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY07 Budget Request. The energy savings models are utilized by the FCVT program for internal project management purposes.

SIngh, M.; Energy Systems; TA Engineering

2008-02-29T23:59:59.000Z

394

Computational analysis of fluid flow and zonal deposition in ferrocyanide single-shell tanks. Ferrocyanide Safety Program  

SciTech Connect

Safety of single-shell tanks containing ferrocyanide wastes is of concern. Ferrocyanide in the presence of an oxidizer such as NaNO{sub 3} or NaNO{sub 2} is explosively combustible when concentrated and heated. Evaluating the processes that could affect the fuel content of waste and distribution of the tank heat load is important. Highly alkaline liquid wastes were transferred in and out of the tanks over several years. Since Na{sub 2}NiFe(CN){sub 6} is much more soluble in alkaline media, the ferrocyanide could be dispersed from the tank more easily. If Cs{sub 2}NiFe(CN){sub 6} or CsNaNiFe(CN){sub 6} are also soluble in alkaline media, solubilization and transport of {sup 137}Cs could also occur. Transporting this heat generating radionuclide to a localized area in the tanks is a potential mechanism for generating a ``hot spot.`` Fluid convection could potentially speed the transport process considerably over aqueous diffusion alone. A stability analysis was performed for a dense fluid layer overlying a porous medium saturated by a less dense fluid with the finding that the configuration is unconditionally unstable and independent of the properties of the porous medium or the magnitude of the fluid density difference. A parametric modeling study of the buoyancy-driven flow due to a thermal gradient was combusted to establish the relationship between the waste physical and thermal properties and natural convection heat transfer. The effects of diffusion and fluid convection on the redistribution of the {sup 137}Cs were evaluated with a 2-D coupled heat and mass transport model. The maximum predicted temperature rise associated with the formation of zones was only 5{degrees}C and thus is of no concern in terms of generating a localized ``hot spot.``

McGrail, B.P.; Trent, D.S.; Terrones, G.; Hudson, J.D.; Michener, T.E.

1993-10-01T23:59:59.000Z

395

Safety Standards  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

US DOE Workshop US DOE Workshop September 19-20, 2012 International perspective on Fukushima accident Miroslav Lipár Head, Operational Safety Section M.Lipar@iaea.org +43 1 2600 22691 2 Content * The IAEA before Fukushima -Severe accidents management * The IAEA actions after Fukushima * The IAEA Action plan on nuclear safety * Measures to improve operational safety * Conclusions THE IAEA BEFORE FUKUSHIMA 4 IAEA Safety Standards IAEA Safety Standards F undamental S afety Principles Safety Fundamentals f o r p ro te c ti n g p e o p l e a n d t h e e n v i ro n m e n t IAEA Safety Standards Regulations for the Safe Transport of Radioactive Material 2005 E dit ion Safety Requirements No. T S-R-1 f o r p ro te c ti n g p e o p l e a n d t h e e n v i ro n m e n t IAEA Safety Standards Design of the Reactor Core for Nuclear Power Plants

396

Safety - Cyclotron  

NLE Websites -- All DOE Office Websites (Extended Search)

Safety The Nuclear Sciences Division (NSD) is committed to providing a safe workplace for its employees, contractors, and guests and conducting its research and operations in a...

397

GenII Gap Analysis  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

GENII-Gap Analysis GENII-Gap Analysis Defense Nuclear Facilities Safety Board Recommendation 2002-1 Software Quality Assurance Improvement Plan Commitment 4.2.1.3: Software Quality Assurance Improvement Plan: GENII Gap Analysis Final Report U.S. Department of Energy Office of Environment, Safety, and Health 1000 Independence Ave., S.W. Washington, DC 20585-2040 May 2004 GENII Gap Analysis May 2004 Final Report INTENTIONALLY BLANK ii GENII Gap Analysis May 2004 Final Report FOREWORD This document provides an evaluation of the Software Quality Assurance (SQA) attributes of GENII, a radiological dispersion computer code, relative to established requirements. This evaluation, a "gap analysis", is performed to meet commitment 4.2.1.3 of the Department of

398

Microsoft Word - Threat Analysis Framework Sept07_comments-final.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5792 5792 Unlimited Release September 2007 Threat Analysis Framework David P. Duggan and John T. Michalski Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy's National Nuclear Security Administration under Contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. Threat Analysis Framework 2 Issued by Sandia National Laboratories, operated for the United States Department of Energy by Sandia Corporation. NOTICE: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government, nor any agency

399

An economic analysis of a monitored retrievable storage site for Tennessee. Final report and appendices  

SciTech Connect

The United States Department of Energy is charged with the task of identifying potential sites for a Monitored Retrievable Storage (MRS) Facility and reporting the results of its analysis to Congress by January 1986. DOE chose three finalist sites from 11 sites DOE analysts evaluated earlier. All three are in Tennessee, including two in Oak Ridge and one in Trousdale/Smith Counties. This paper is a summary of research undertaken on the economic effects of establishing the MRS facility in Tennessee. All three locations were considered in the analysis, but on some occasions attention is focused on the site preferred by DOE. The research was undertaken by the Center for Business and Economic Research (CBER), College of Business Administration, the University of Tennessee, Knoxville, under contract with the Tennessee Department of Economic and Community Development.

Fox, W.F.; Mayo, J.W.; Hansen, L.T.; Quindry, K.E.

1985-12-17T23:59:59.000Z

400

Microsoft Word - ORO ISC Functional Analysis and Inventory 3-6-07 FINAL.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

Table of Contents Page Introduction 1 Mission 1 Distinctive Characteristics 2 Assumptions 3 Staffing and Trends 4 Critical Skills Inventory 6 Significant Successes 7 Strategic Goals and Objectives 9 Conclusions 10 Appendix A: Functional Descriptions 11 Appendix B: SC ISC Staffing Levels by Occupational Groupings 16 Appendix C: Organizational Metrics 21 F U N C T I O N A L A N A L Y S I S A N D I N V E N T O R Y Introduction The Manager, Oak Ridge Office (ORO), has requested this Functional Analysis and Inventory to identify and present quantifiable performance metrics for various matrix support activities performed by ORO. This analysis will be used to describe the ORO operating model as a component of the Office of Science (SC) Integrated

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Microsoft Word - Lamellae tungsten tile design thermal and electromagnetic stress analysis_Final.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

Lamellae tungsten tile design transient Lamellae tungsten tile design transient thermal/ electromagnetic stress analysis Thomas Willard*, Rui Vieira, Samuel Pierson MIT Plasma Science and Fusion Center, Cambridge, MA 02139 8 June 2006 Abstract A transient thermal/ electromagnetic stress analysis of the lamellae tungsten tile design has been performed to determine if the design is adequate to meet the maximum design load conditions of 12 MW/ m 2 uniform heat flux for 5 seconds (single pulse, no Diverter Plate temperature ratcheting) , superimposed on the electromagnetic body load due to eddy currents generated by disruptions. The results show that the design is adequate, with the stresses in the tungsten lamellae and the TZM molybdenum hardware less than the ultimate strength of

402

Benefit-cost analysis of DOE's Current Federal Program to increase hydrothermal resource utilization. Final report  

DOE Green Energy (OSTI)

The impact of DOE's Current Federal Program on the commercialization of hydrothermal resources between 1980 and 2000 is analyzed. The hydrothermal resources of the United States and the types of DOE activities used to stimulate the development of these resources for both electric power and direct heat use are described briefly. The No Federal Program and the Current Federal Program are then described in terms of funding levels and the resultant market penetration estimates through 2000. These market penetration estimates are also compared to other geothermal utilization forecasts. The direct benefits of the Current Federal Program are next presented for electric power and direct heat use applications. An analysis of the external impacts associated with the additional hydrothermal resource development resulting from the Current Federal Program is also provided. Included are environmental effects, national security/balance-of-payments improvements, socioeconomic impacts and materials requirements. A summary of the analysis integrating the direct benefits, external impacts and DOE program costs concludes the report.

Not Available

1981-12-10T23:59:59.000Z

403

Application analysis of solar total energy systems to the residential sector. Volume IV, market penetration. Final report  

DOE Green Energy (OSTI)

This volume first describes the residential consumption of energy in each of the 11 STES regions by fuel type and end-use category. The current and projected costs and availability of fossil fuels and electricity for the STES regions are reported. Projections are made concerning residential building construction and the potential market for residential STES. The effects of STES ownership options, institutional constraints, and possible government actions on market penetration potential were considered. Capital costs for two types of STES were determined, those based on organic Rankine cycle (ORC) heat engines and those based on flat plate, water-cooled photovoltaic arrays. Both types of systems utilized parabolic trough collectors. The capital cost differential between conventional and STE systems was calculated on an incremental cost per dwelling unit for comparison with projected fuel savings in the market penetration analysis. The market penetration analysis was planned in two phases, a preliminary analysis of each of the geographical regions for each of the STE systems considered; and a final, more precise analysis of those regions and systems showing promise of significant market penetration. However, the preliminary analysis revealed no geographical regions in which any of the STES considered promised to be competitive with conventional energy systems using utility services at the prices projected for future energy supplies in the residential market. Because no promising situations were found, the analysis was directed toward an examination of the parameters involved in an effort to identify those factors which make a residential STES less attractive than similar systems in the commercial and industrial areas. Results are reported. (WHK)

Not Available

1979-07-01T23:59:59.000Z

404

INVESTIGATION OF MAGNETOHYDRODYNAMIC POWER GENERATION. VOLUME I. SYSTEM ANALYSIS AND ENVIRONMENTAL TESTING. Final Report  

SciTech Connect

A system analysis was performed to establish the design and performance characteristics of 250- and 1000-Mev MHD generators. The results are presented, and the necessary equations are derived. The results of environmental tests that investigated boundary and bulk conductivity, structural concepts, and combustion phenomena are included. The other five volumes of the report are summarized, together with the principal conclusions. (D.C.W.)

1963-05-01T23:59:59.000Z

405

Benefits analysis for the production of fuels and chemicals using solar thermal energy. Final report  

DOE Green Energy (OSTI)

Numerous possibilities exist for using high temperature solar thermal energy in the production of various chemicals and fuels (Sun Fuels). Research and development activities have focused on the use of feedstocks such as coal and biomass to provide synthesis gas, hydrogen, and a variety of other end-products. A Decision Analysis technique geared to the analysis of Sun Fuels options was developed. Conventional scoring methods were combined with multi-attribute utility analysis in a new approach called the Multi-Attribute Preference Scoring (MAPS) system. MAPS calls for the designation of major categories of attributes which describe critical elements of concern for the processes being examined. The six major categories include: Process Demonstration; Full-Scale Process, Feedstock; End-Product Market; National/Social Considerations; and Economics. MAPS calls for each attribute to be weighted on a simple scale for all of the candidate processes. Next, a weight is assigned to each attribute, thus creating a multiplier to be used with each individual value to derive a comparative weighting. Last, each of the categories of attributes themselves are weighted, thus creating another multiplier, for use in developing an overall score. With sufficient information and industry input, each process can be ultimately compared using a single figure of merit. After careful examination of available information, it was decided that only six of the 20 candidate processes were adequately described to allow a complete MAPS analysis which would allow direct comparisons for illustrative purposes. These six processes include three synthesis gas processes, two hydrogen and one ammonia. The remaining fourteen processes were subjected to only a partial MAPS assessment.

None

1982-05-01T23:59:59.000Z

406

An Energy and Peak Loads Analysis of the TYC/TRC Building Final Report  

E-Print Network (OSTI)

The energy use of the Texas Youth Commission/Texas Rehabilitation Commission (TYC/TRC) Building at Austin, Texas, was analyzed using the DOE 2.IB building energy simulation program. An analysis was made for the building as specified in the building plans and the specifications provided by the State Purchasing and General Services Commission. Operating schedules for occupancy, lighting, office equipment, and infiltration were assumed. The energy consumption of the TYC/TRC Building can be reduced with certain modifications.

Katipamula, S.; O'Neal, D. L.

1987-01-01T23:59:59.000Z

407

Analysis of the need for intermediate and peaking technologies in the year 2000. Final report  

SciTech Connect

This analysis was conducted to assess the impact of load management on the future need for intermediate- and peak-generating technologies (IPTs) such as combustion turbines, pumped storage, and cycling coal plants. There would be a reduced need for IPTs if load-management activities such as time-of-use pricing, together with customer-owned energy-storage devices, hot-water-heater controls, and interruptible service can economically remove most of the variation from electric power demands. The objective of this analysis is to assess the need for IPTs in an uncertain future, which will probably include load management and time-differentiated electricity prices. The analysis is exploratory in nature and broad in scope. It does not attempt to predict the future or to model precisely the technical characteristics or economic desirability of load management. Rather, its purpose is to provide research and development planners with some basic insights into the order of magnitude of possible hourly demand shifts on a regional basis and to determine the impact of load management on daily and seasonal variations in electricity demand.

Barrager, S.M.; Campbell, G.L.

1980-04-01T23:59:59.000Z

408

Photovoltaic power systems market identification and analysis. Final report, January 1977--February 1978  

DOE Green Energy (OSTI)

This report summarizes the work done by InterTechnology/Solar Corporation, its consultants, Mobil Tyco Solar Energy Corporation and the University of Delaware Institute for Energy Conversion, and its consultants, during the marketing analysis of near and intermediate term photovoltaic power applications. To obtain estimates of the domestic and foreign market potential for photovoltaically powered devices two approaches were used. First, the study was identifying then screening all possible photovoltaic power supply applications. This approach encompassed the first two tasks of the study: (1) a survey of the current uses of photovoltaic systems, both domestic and international, and a projection of the usage of those systems into the future; and (2) a new idea generation task which attempted to come up with new ways of using photovoltaic power. Second, the study required in-depth analysis of key near-term and intermediate-term photovoltaic applications identified during the first phase to obtain reasonable estimates of photovoltaic market potential. This process encompassed the third and fourth tasks of the analysis: (3) refinement of ideas generated in Task 2 so that certain products/applications could be identified, the product defined and a market survey carried out; and (4) development of a detailed product scenario which forecasts sales, barriers to market acceptance, and technical innovationsrequired for proper introduction of the products. The work performed and findings of each task are presented.

Not Available

1979-05-01T23:59:59.000Z

409

Liquid chromatographic analysis of coal surface properties. Final report, September 1991--February 1995  

SciTech Connect

Experiments on equilibrium adsorption loadings of various probe compounds on 60-200 mesh Illinois {number_sign}6 coal (PSOC-1539), Adaville {number_sign}1 coal (PSOC-1544), Wyodak coal (PSOC-1545) and Pittsburgh {number_sign}8 coal (PSOC-1549) were performed. the probe compounds include m-cresol, p-cresol, o-cresol, phenol, n-octanol, n-heptanol, n-propanol, isopropanol n-butanol, s-butanol, 2-butanol, t-butanol, 2-naphthol, cyclohexanol, 2-methyl-1-pentanol (2M1P), 4-methyl-2-pentanol (4M2P), benzene and toluene. Equilibrium adsorption of various probe compounds on the coals were measured with the inverse liquid chromatography method. Experiments on flotation of various 60-200 mesh treated coals such as Illinois {number_sign}6 coal (PSOC-1539), Adaville {number_sign}1 coal (PSOC-1544), Wyodak coal (PSOC-1545) and Pittsburgh {number_sign}8 coal (PSOC-1549) were performed. The chosen coals were treated with steam, nitrogen and air at 1 atm and 125-225{degrees}C for 24 hours. The coals were treated with water as well as 20-1000 ppm aqueous alcohol solutions for 3-24 hours at 150-225{degrees}C. The coals also were treated with 20-ppm alcohol aqueous solutions for 1-24 hours at the 0.002-g/min mass flow rate of alcohol aqueous solutions and at 225{degrees}C. Flotation experiments were conducted with a 500-cm{sup 3} batch-type micro flotation apparatus, introducing nitrogen at the bottom of the apparatus. This final report was prepared with the experimental data obtained during the period of September 1991-March 1994.

Kwon, K.C.

1996-03-01T23:59:59.000Z

410

Reduced order models for thermal analysis : final report : LDRD Project No. 137807.  

SciTech Connect

This LDRD Senior's Council Project is focused on the development, implementation and evaluation of Reduced Order Models (ROM) for application in the thermal analysis of complex engineering problems. Two basic approaches to developing a ROM for combined thermal conduction and enclosure radiation problems are considered. As a prerequisite to a ROM a fully coupled solution method for conduction/radiation models is required; a parallel implementation is explored for this class of problems. High-fidelity models of large, complex systems are now used routinely to verify design and performance. However, there are applications where the high-fidelity model is too large to be used repetitively in a design mode. One such application is the design of a control system that oversees the functioning of the complex, high-fidelity model. Examples include control systems for manufacturing processes such as brazing and annealing furnaces as well as control systems for the thermal management of optical systems. A reduced order model (ROM) seeks to reduce the number of degrees of freedom needed to represent the overall behavior of the large system without a significant loss in accuracy. The reduction in the number of degrees of freedom of the ROM leads to immediate increases in computational efficiency and allows many design parameters and perturbations to be quickly and effectively evaluated. Reduced order models are routinely used in solid mechanics where techniques such as modal analysis have reached a high state of refinement. Similar techniques have recently been applied in standard thermal conduction problems e.g. though the general use of ROM for heat transfer is not yet widespread. One major difficulty with the development of ROM for general thermal analysis is the need to include the very nonlinear effects of enclosure radiation in many applications. Many ROM methods have considered only linear or mildly nonlinear problems. In the present study a reduced order model is considered for application to the combined problem of thermal conduction and enclosure radiation. The main objective is to develop a procedure that can be implemented in an existing thermal analysis code. The main analysis objective is to allow thermal controller software to be used in the design of a control system for a large optical system that resides with a complex radiation dominated enclosure. In the remainder of this section a brief outline of ROM methods is provided. The following chapter describes the fully coupled conduction/radiation method that is required prior to considering a ROM approach. Considerable effort was expended to implement and test the combined solution method; the ROM project ended shortly after the completion of this milestone and thus the ROM results are incomplete. The report concludes with some observations and recommendations.

Hogan, Roy E., Jr.; Gartling, David K.

2010-09-01T23:59:59.000Z

411

Conceptual design and systems analysis of photovoltaic power systems. Volume II. Systems. Revised final report  

DOE Green Energy (OSTI)

Conceptual designs were made and analyses were performed on three types of solar photovoltaic power systems. Included were Residential (1--10 kW), Intermediate (0.1--10 MW), and Central (50--1000 MW) Power Systems to be installed in the 1985 to 2000 time period. Detailed descriptions of each of the three systems studied, descriptions of the necessary subsystems, and discussions of the interfaces between them are presented. Included also are descriptions of system performance and system cost used to perform an economic analysis which assesses the value of each system.

Pittman, P.F.

1977-03-01T23:59:59.000Z

412

Conceptual design and systems analysis of photovoltaic power systems. Final report. Volume III(2). Technology  

DOE Green Energy (OSTI)

Conceptual designs were made and analyses were performed on three types of solar photovoltaic power systems. Included were Residential (1 to 10 kW), Intermediate (0.1 to 10 MW), and Central (50 to 1000 MW) Power Systems to be installed in the 1985 to 2000 time period. The following analyses and simulations are covered: residential power system computer simulations, intermediate power systems computer simulation, central power systems computer simulation, array comparative performance, utility economic and margin analyses, and financial analysis methodology.

Pittman, P.F.

1977-05-01T23:59:59.000Z

413

Multiattribute decision analysis method for evaluating buildings and building systems. Final report  

SciTech Connect

Multiattribute decision analysis (MADA) methods consider non-financial attributes (qualitative and quantitative) in addition to common financial worth measures when evaluating project alternatives. The report reviews 14 classes of methods for performing MADA. It summarizes their usefulness for screening, ranking, and choosing among projects; their data input requirements; and how each method scores project alternatives. Two methods--the analytical hierarchy process (AHP) and non-traditional capital investment criteria (NCIC)--are described in detail. Assumptions, procedures, strengths, and limitations are described for each.

Norris, G.A.; Marshall, H.E.

1995-09-01T23:59:59.000Z

414

Analysis of Chinook Salmon in the Columbia River from an Ecosystem Perspective. Final Report.  

DOE Green Energy (OSTI)

Ecosystem Diagnosis and Treatment (EDT) methodology was applied to the analysis of chinook salmon in the mid-Columbia subbasins which flow through the steppe and steppe-shrub vegetation zones. The EDT examines historical changes in life history diversity related to changes in habitat. The emphasis on life history, habitat and historical context is consistent with and ecosystem perspective. This study is based on the working hypothesis that the decline in chinook salmon was at least in part due to a loss of biodiversity defined as the intrapopulation life history diversity. The mid Columbia subbasins included in the study are the Deschutes, John Day, Umatilla, Tucannon and Yakima.

Lichatowich, James A.; Mobrand, Lars E.

1995-01-01T23:59:59.000Z

415

Economic Analysis of the Environmental Effects of the Coal-Fired Electric Generator at Boardman, Oregon. Final Report.  

SciTech Connect

This study is one of several commissioned by the Bonneville Power Administration (BPA) to estimate the economic value of the environmental costs and benefits of different electricity-generating resources. In it we described and quantify the environmental costs and benefits of coal-fired generators, using the plant in Boardman, Oregon, as the basis for our estimations. The Boardman plant uses pulverized coal to produce steam for generating electricity. It is nominally rated at 550 megawatts. This study assumes a 70% load factor and an annual production of 3373 x 10/sup 6/ kWh. Cooling water comes from a 1400-acre cooling pond; coal comes from Wyoming in 100-car unit-trains every two days. The estimated service life of the plant is 40 years. We developed a socioeconomic-environmental model to assess the final physical impacts of each of the initial impacts resulting from the fuel cycle. The analysis