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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.


1

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

2

Fuel Storage Facility Final Safety Analysis Report. Revision 1  

SciTech Connect

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

Linderoth, C.E.

1984-03-01T23:59:59.000Z

3

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

SciTech Connect

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

4

Fast Flux Test Facility final safety analysis report. Amendment 73  

SciTech Connect

This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

Gantt, D.A.

1993-08-01T23:59:59.000Z

5

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

SciTech Connect

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

Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

1990-11-09T23:59:59.000Z

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

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

8

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

9

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

10

CACI: The Cesium-137 Agricultural Commodities Irradiator. Final design report: Volume 7, Safety analysis, thermal analysis, and thermal testing  

SciTech Connect

This report provides a complete description of the final detailed design of the Cesium-137 Agricultural Commodities Irradiator (CACI). The design was developed and successfully completed by the Rocketdyne Division of Rockwell International for the US Department of Energy (DOE). The CACI project was initiated in April 1985 under DOE`s Byproducts Utilization Program, with the objectives of transferring food irradiation technology to the industry and thereby demonstrating a beneficial use for the 137 Cs nuclear by-product isotope. As designed, CACI will meet the intended requirements for research, development, and demonstration of irradiation processing of food. Further, as shown in the safety analyses performed during the project, the design conforms to all the safety and licensing requirements set forth for the project. The original scope of the CACI project included completion of its construction. However, the project was terminated for the convenience of the government during the final design phase in February 1986 for lack of a specific site. The CACI final design is described in eight volumes. This volume, Volume VII, describes Safety Analysis, Thermal Analysis, and Thermal Testing.

Not Available

1986-12-19T23:59:59.000Z

11

Final report for confinement vessel analysis. Task 2, Safety vessel impact analyses  

SciTech Connect

This report describes two sets of finite element analyses performed under Task 2 of the Confinement Vessel Analysis Program. In each set of analyses, a charge is assumed to have detonated inside the confinement vessel, causing the confinement vessel to fail in either of two ways; locally around the weld line of a nozzle, or catastrophically into two hemispheres. High pressure gases from the internal detonation pressurize the inside of the safety vessel and accelerate the fractured nozzle or hemisphere into the safety vessel. The first set of analyses examines the structural integrity of the safety vessel when impacted by the fractured nozzle. The objective of these calculations is to determine if the high strength bolt heads attached to the nozzle penetrate or fracture the lower strength safety vessel, thus allowing gaseous detonation products to escape to the atmosphere. The two dimensional analyses predict partial penetration of the safety vessel beneath the tip of the penetrator. The analyses also predict maximum principal strains in the safety vessel which exceed the measured ultimate strain of steel. The second set of analyses examines the containment capability of the safety vessel closure when impacted by half a confinement vessel (hemisphere). The predicted response is the formation of a 0.6-inch gap, caused by relative sliding and separation between the two halves of the safety vessel. Additional analyses with closure designs that prevent the gap formation are recommended.

Murray, Y.D. [APTEK, Inc., Colorado Springs, CO (United States)

1994-01-26T23:59:59.000Z

12

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

SciTech Connect

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

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

1993-09-01T23:59:59.000Z

13

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

SciTech Connect

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

14

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

SciTech Connect

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

15

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

16

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.

17

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

Office of Environmental Management (EM)

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

18

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

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

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

19

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

20

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)

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

Ferrocyanide safety project ferrocyanide aging studies. Final report  

SciTech Connect

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

22

CRAD, Facility Safety- Documented Safety Analysis  

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

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

23

Events Beyond Design Safety Basis Analysis | Department of Energy  

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

Events Beyond Design Safety Basis Analysis Events Beyond Design Safety Basis Analysis March 23, 2011 Safety Bulletin 2011-01, Events Beyond Design Safety Basis Analysis This Safety...

24

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

25

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

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

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

26

B PLANT DOCUMENTED SAFETY ANALYSIS  

SciTech Connect

This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the 221-B Canyon Building and ancillary support structures (B Plant). The document replaces BHI-010582, Documented Safety Analysis for the B-Plant Facility. The B Plant is non-operational, deactivated and undergoing long term S&M prior to decontamination and decommissioning (D&D). This DSA is compliant with 10 CFR 830, Nuclear Safety Management, Subpart B, ''Safety Basis Requirements.'' The DSA was developed in accordance with U.S. Department of Energy (DOE) standard DOE-STD-1120-98, Integration of Environment, Safety, and Health into Facility Disposition Activities (DOE 1998) per Table 2 of 10 CFR 830 Appendix A, DOE Richland Operation Office (RL) direction (02-ABD-0053, Fluor Hanford Nuclear Safety Basis Strategy and Criteria) for facilities in long term S&M, and RL Direction (02-ABD-0091, ''FHI Nuclear Safety Expectations for Nuclear Facilities in Surveillance and Maintenance''). A crosswalk was prepared to identify potential inconsistencies between the previous B Plant safety analysis and DOE-STD-1120-98 guidance. In general, the safety analysis met the criteria of DOE-STD-1120-98. Some format and content changes have been made, including incorporating recent facility modifications and updating the evaluation guidelines and control selection criteria in accordance with RL direction (02-ABD-0053). The facility fire hazard analysis (FHA) and Technical Safety Requirements (TSR) are appended to this DSA as an aid to the users, to minimize editorial redundancy, and to provide an efficient basis for update.

DODD, E.N.; KERR, N.R.

2003-08-01T23:59:59.000Z

27

Waste Isolation Pilot Plant Safety Analysis Report  

SciTech Connect

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

28

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:

29

Improving the safety of LWR power plants. Final report  

SciTech Connect

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

Not Available

1980-04-01T23:59:59.000Z

30

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

31

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

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

Analysis of DOE Safety Performance Current Safety Performance Trends The Office of Environment, Health, Safety and Security, Office of Analysis provides analysis of Department of...

32

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

33

Monthly Analysis of Electrical Safety Occurrences - February...  

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

2 Monthly Analysis of Electrical Safety Occurrences - February 2012 February 2012 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by...

34

Monthly Analysis of Electrical Safety Occurrences - November...  

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

1 Monthly Analysis of Electrical Safety Occurrences - November 2011 November 2011 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by...

35

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

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

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

36

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

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

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

37

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

38

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

Energy Savers (EERE)

Monthly Analysis of Electrical Safety Occurrences - March 2011 Monthly Analysis of Electrical Safety Occurrences - March 2011 March 2011 An analysis of the Occurrence Reporting and...

39

Supplemental Analysis for the Final Environmental Impact Statement...  

Office of Environmental Management (EM)

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

40

EA-1212-SA-01: Final Supplement Analysis  

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

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

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

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

42

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

Energy Savers (EERE)

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...

43

FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND...  

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

HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE November 14, 2013 Richland, WA Topics in this Meeting Summary Opening ......

44

Motorway Safety in Europe and Greece: A Comparative Analysis  

Science Journals Connector (OSTI)

Motorways offer a high level of road safety as compared to other road types. In most EU member states, motorways represent a percentage of all primary and secondary roads varying between 1% and 12% (EU15). In Greece this figure has been significantly lower until 2004 and since then it has been sharply increased. Road safety levels have been also improved as a consequence of that. The paper provides a comparative analysis of motorway safety among Greece and the EU and furthermore among Greek motorways. To achieve this, it makes use of certain safety indicators such as fatality and accident rates. In addition, for the Greek motorways, it makes use of additional data which are available from the motorway operators data bases including incidents and their causes. Association of accidents to drivers behavior is sought where possible. Finally, a more in depth investigation takes place concerning accidents in the Greek motorways, mainly in the ATTICA TOLLWAY and in EGNATIA ODOS.

Dimitrios Papaioannou; Alexandros Kokkalis

2012-01-01T23:59:59.000Z

45

Autoclave nuclear criticality safety analysis  

SciTech Connect

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

D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

1991-12-31T23:59:59.000Z

46

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

47

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.

48

On the Integration of Requirements Analysis and Safety Analysis for Safety-Critical Software  

E-Print Network (OSTI)

this is acceptable within the context of system risk. The advantage of conducting the safety analysis during is acceptable within the overall #12;2 system risk. If the risk is not acceptable the safety specification has analysis is to determine the risk associated with requirements specifications and assess whether

Newcastle upon Tyne, University of

49

K West integrated water treatment system subproject safety analysis document  

SciTech Connect

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

50

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

51

Final Report on the Safety Assessment of Aluminum Silicate, Calcium Silicate, Magnesium Aluminum  

E-Print Network (OSTI)

Final Report on the Safety Assessment of Aluminum Silicate, Calcium Silicate, Magnesium Aluminum Silicate, Magnesium Silicate, Magnesium Trisilicate, Sodium Magnesium Silicate, Zirconium Silicate, Attapulgite, Bentonite, Fuller's Earth, Hectorite, Kaolin, Lithium Magnesium Silicate, Lithium Magnesium

Ahmad, Sajjad

52

Hanford safety analysis and risk assessment handbook (SARAH)  

SciTech Connect

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

GARVIN, L.J.

2003-01-20T23:59:59.000Z

53

DOE high-level waste tank safety program. Final report  

SciTech Connect

The overall objective of the work was to provide LANL with support to the DOE High-Level Waste Tank Safety Program. This effort included direct support to the DOE High-Level Waste Tank Working Groups, development of a database to track all identified safety issues, development of requirements for waste tank modernization, evaluation of external comments regarding safety-related guidance/instruction developed previously, examination of current federal and state regulations associated with DOE Tank farm operations, and performance of a conduct of operations review. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided.

NONE

1998-11-01T23:59:59.000Z

54

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

55

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

56

SAFETY AND RELIABILITY ANALYSIS OF NUCLEAR REACTORS  

Science Journals Connector (OSTI)

Abstract A survey of the various aspects of safety and reliability analysis of nuclear reactors is presented with particular emphasis on the interrelation between structural reliability and systems reliability. In reactor design this interrelation is of overriding importance since it is the task of the control, protective and containment systems to protect the mechanical system and the structure from accidental overloading.

T.A. JAEGER

1972-01-01T23:59:59.000Z

57

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

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

impact of the review team on the quality of the final documents. The reports also document the tracking and closure of a number of conditions of approval or directions...

58

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

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

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

59

Microsoft PowerPoint - Module 3 - Safety Design Approach - final...  

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

and more straight-forward engineering analysis 4 engineering analysis. NRC Advanced Reactor Policy Statement (22) * Designs which minimize the potential for severe accidents...

60

On March 26, 2012, the Safety Culture Task Force (SCTF) of the Committee on Chemical Safety, American Chemical Society, published the final draft of its report on  

E-Print Network (OSTI)

safety. Implement hazards analysis procedures in all new lab work, especially laboratory research. (The procedures in all new lab work, especially laboratory research. 7. Build awareness and caring for safety for safety. 9. Adopt a personal credo: the "Safety Ethic"--value safety, work safely, prevent at

Farritor, Shane

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

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

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

Industrial Glass Bandwidth Analysis Final Report, August 2007 ITP Glass: Industrial Glass Bandwidth Analysis Final Report, August 2007 industrialbandwidth.pdf More Documents &...

62

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

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

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

63

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.

64

Regulatory analysis technical evaluation handbook. Final report  

SciTech Connect

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

65

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.

66

242-A evaporator safety analysis report  

SciTech Connect

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

CAMPBELL, T.A.

1999-05-17T23:59:59.000Z

67

Safety analysis of in-use vehicle wrapping cylinder  

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

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

68

Monthly Analysis of Electrical Safety Occurrences April 2013  

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

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

69

Monthly Analysis of Electrical Safety Occurrences March 2013  

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

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

70

Monthly Analysis of Electrical Safety Occurrences August 2013  

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

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

71

Monthly Analysis of Electrical Safety Occurrences November 2012  

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

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

72

Monthly Analysis of Electrical Safety Occurrences June 2012  

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

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

73

Monthly Analysis of Electrical Safety Occurrences April 2012  

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

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

74

Monthly Analysis of Electrical Safety Occurrences December 2012  

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

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

75

Monthly Analysis of Electrical Safety Occurrences July 2013  

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

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

76

Monthly Analysis of Electrical Safety Occurrences August 2012  

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

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

77

Monthly Analysis of Electrical Safety Occurrences May 2012  

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

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

78

Monthly Analysis of Electrical Safety Occurrences October 2012  

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

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

79

Monthly Analysis of Electrical Safety Occurrences September 2011  

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

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

80

Monthly Analysis of Electrical Safety Occurrences October 2011  

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

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

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

Monthly Analysis of Electrical Safety Occurrences September 2013  

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

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

82

Monthly Analysis of Electrical Safety Occurrences February 2013  

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

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

83

Monthly Analysis of Electrical Safety Occurrences June 2011  

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

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

84

Monthly Analysis of Electrical Safety Occurrences May 2013  

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

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

85

Monthly Analysis of Electrical Safety Occurrences December 2011  

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

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

86

Monthly Analysis of Electrical Safety Occurrences August 2011  

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

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

87

Monthly Analysis of Electrical Safety Occurrences March 2012  

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

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

88

Monthly Analysis of Electrical Safety Occurrences July 2011  

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

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

89

Monthly Analysis of Electrical Safety Occurrences September 2012  

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

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

90

Monthly Analysis of Electrical Safety Occurrences January 2013  

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

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

91

Monthly Analysis of Electrical Safety Occurrences July 2012  

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

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

92

Monthly Analysis of Electrical Safety Occurrences June 2013  

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

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

93

Monthly Analysis of Electrical Safety Occurrences January 2012  

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

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

94

Comparison of Intergrated Safety Analysis (ISA) and Probabilistic...  

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

Commission Washington, DC 20555-0001 SUBJECT: COMPARISON OF INTEGRATED SAFETY ANALYSIS (ISA) AND PROBABILISTIC RISK ASSESSMENT (PRA) FOR FUEL CYCLE FACILITIES Dear Chairman...

95

CRAD, Documented Safety Analysis Development- April 23, 2013  

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

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

96

Radiation safety considerations for the parasitic Final Focus Test Beam at SLAC  

SciTech Connect

A low intensity electron beam parasitic to the operation of the Stanford Linear Collider (SLC) has been transported through the Final Focus Test Beam (FFTB) facility making secondary test beams available for users. Photons generated in collimation of the SLC electron and positron beams in the linac pass through a splitter magnet that deflects the primary beams away from the linac axis into the SLC beam lines. These photons are converted to electrons and positrons in a secondary production target located down beam on the linac axis. The secondary electrons are then transported through the FFTB beam line onto experimental detectors. The average power of the parasitic beam is very low, thus, it presents no hazards. However, various accident scenarios involving failure of the splitter magnet and the active protection devices could send much more powerful SLC beams (up to 90 kilo-watts) into this zero-degree secondary beam line. For the accident cases, the average power in the transmitted beam was calculated using the Monte Carlo programs EGS4 and TURTLE. Results from analysis of the radiation protection systems that assure safety during the parasitic operation are presented.

Rokni, S.H.; Iverson, R.H.; Keller, L.P.

1996-11-01T23:59:59.000Z

97

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

98

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

99

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

100

DOE Hydrogen Transition Analysis Workshop: Final Agenda | Department...  

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

Agenda DOE Hydrogen Transition Analysis Workshop: Final Agenda Agenda for the DOE Hydrogen Transition Analysis Workshop on January 26, 2006. transitionwkshpagenda.pdf More...

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 ENVIRONMENTAL ASSESSMENT / FINAL REGULATORY IMPACT REVIEW / INITIAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network (OSTI)

FINAL ENVIRONMENTAL ASSESSMENT / FINAL REGULATORY IMPACT REVIEW / INITIAL REGULATORY FLEXIBILITY ..............................................................................................16 2 REGULATORY IMPACT REVIEW ANALYSIS for Amendment 83 to the Fishery Management Plan for Groundfish of the Gulf of Alaska ALLOCATION

102

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

SciTech Connect

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

103

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

104

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems  

E-Print Network (OSTI)

Tree Analysis), FMEA (Failure Mode and Effect Analysis), HAZOP (Hazard and Operability study). · Safety

105

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

106

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

107

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect

Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 26 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. In one of the loss-of-flow accidents offsite electrical power is assumed lost to the three operating primary pumps. A slightly delayed reactor scram is initiated as a result of primary flow coast down. The RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and that the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail, shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail. An additional analysis was performed to simulate the operation at low power (500 kW) without forced flow cooling. The result indicates that natural convection cooling is adequate for operation of the NBSR at a power level of 500 kW.

CHENG,L.HANSON,A.DIAMOND,D.XU,J.CAREW,J.RORER,D.

2004-03-31T23:59:59.000Z

108

Software FMEA analysis for safety-related application software  

Science Journals Connector (OSTI)

Abstract A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests.

Gee-Yong Park; Dong Hoon Kim; Dong Young Lee

2014-01-01T23:59:59.000Z

109

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

SciTech Connect

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

110

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

111

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......System for Research Reactor (IRSRR). Available...System for Research Reactor (IRSRR). Available...76. 7 Manual on reliability data collection for research reactor PSAs. (1992) IAEA...probabilistic safety analysis of incidents in nuclear......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

112

Helicopter final assembly critical path analysis  

E-Print Network (OSTI)

Helicopter final assembly involves the installation of hundreds of components into the aircraft and takes thousands of man-hours. Meeting production targets such as total build days and total aircraft man-hours can be ...

Daigh, Sara L. (Sarah Louise), 1981-

2012-01-01T23:59:59.000Z

113

Hazard Analysis Database report  

SciTech Connect

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

114

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

115

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

116

Software safety analysis of function block diagrams using fault trees , Junbeom Yoob,*, Sungdeok Chab  

E-Print Network (OSTI)

requirements written in natural language and performing safety analysis techniques such as FMEA [5] and HAZOP

117

Safety analysis for Shippingport Station Decommissioning Project. Vol. 9. Pt. 1. Rev. 1  

SciTech Connect

Information is presented concerning the safety analysis for the decommissioning project; and permitting plan.

Not Available

1983-09-01T23:59:59.000Z

118

The conservation planning analysis model. Final report  

SciTech Connect

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

Not Available

1993-12-31T23:59:59.000Z

119

EIS-0350-SA-02: Final Supplement Analysis  

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

Chemistry and Metallurgy Research Building Replacement Project at Los Alamos National Laboratory, Los Alamos, New Mexico Supplement Analysis

120

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

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

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

122

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect

Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 30 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. Two cases were considered for loss of electrical power. In the first case offsite power is lost, resulting in an immediate scram caused by loss of power to the control rod system. In the second case power is lost to only the three operating primary pumps, resulting in a slightly delayed scram when loss-of-flow is detected as the pumps coast down. In both instances, RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail.

CAREW,J.CHENG,L.HANSON,AXU,J.RORER,D.DIAMOND,D.

2003-08-26T23:59:59.000Z

123

A new DOE standard for transuranic waste nuclear safety analysis  

SciTech Connect

The DOE Office of Environmental Management (EM) observed through onsite assessments and a review of site-specific lessons learned that transuranic (TRU) waste operations could benefit from standardization of assumptions and approaches used to analyze hazards and select controls. EM collected and compared safety analysis information from DOE sites, including a comparison of the type of TRU waste accidents evaluated and controls selected, as well as specific Airborne Release Fractions (ARFs), Respirable Fractions (RFs), and Damage Ratios (DRs) assumed in accident analyses. This paper recounts the efforts by the DOE and its contractors to bring consistency to the safety analysis process supporting TRU waste operations through an integrated re-engineering effort. EM embarked on a process to re-engineer and standardize TRU safety analysis activities complex-wide. The effort involved DOE headquarters, field offices, and contractors. Five teams were formed to analyze and develop the necessary technical basis for a DOE Technical Standard. The teams looked at general issues including Safety Basis (SB), drum integrity and inspection criteria, hazard controls and analysis, safety analysis review and approval process, and implementation of hazard controls. (authors)

Triay, I.; Chung, D. [U.S. Department of Energy, Washington, D.C. (United States); Woody, J. [Atlas Consulting, Knoxville, TN (United States); Foppe, T. [Carlsbad Technical Assistance Contractor, Carlsbad, NM (United States); Mewhinney, C. [Sandia National Laboratories, Carlsbad, NM (United States); Jennings, S. [Los Alamos National Laboratories, Carlsbad, NM (United States)

2007-07-01T23:59:59.000Z

124

Final Report- Hydrogen Delivery Infrastructure Options Analysis  

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

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

125

Sandis irradiator for dried sewage solids. Final safety analysis report  

SciTech Connect

Analyses of the hazards associated with the operation of the Sandia irradiator for dried sewage solids, as well as methods and design considerations to minimize these hazards, are presented in accordance with DOE directives.

Morris, M.

1980-07-01T23:59:59.000Z

126

RISMC ADVANCED SAFETY ANALYSIS WORKING PLAN FY 2015 FY 2019  

SciTech Connect

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

Ronaldo H. Szilard; Curtis L. Smith

2014-09-01T23:59:59.000Z

127

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

128

System safety analysis of an autonomous mobile robot  

SciTech Connect

Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate{trademark} robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA{copyright}) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection.

Bartos, R.J.

1994-08-01T23:59:59.000Z

129

FAQS Gap Analysis Qualification Card Occupational Safety  

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

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

130

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

131

ENVIRONMENTAL ASSESSMENT / REGULATORY IMPACT REVIEW / FINAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network (OSTI)

ENVIRONMENTAL ASSESSMENT / REGULATORY IMPACT REVIEW / FINAL REGULATORY FLEXIBILITY ANALYSIS for Modifying existing Chinook and chum salmon savings areas AMENDMENT 84 to the Fishery Management Plan by the current regulatory closure regulations, as much higher salmon bycatch rates are reportedly encountered

132

ALTERNATIVE JET FUEL SCENARIO ANALYSIS Final Report  

E-Print Network (OSTI)

and considers existing and emerging fuel production technologies. The analysis also forecasts how alternative fuels might contribute to greenhouse gas goals. Based on a review of fuel production companies' stated of the most optimistic demand forecasts and the "product switch" production scenarios leads to North American

133

Final report on the Pathway Analysis Task  

SciTech Connect

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

Whicker, F.W.; Kirchner, T.B. [Colorado State Univ., Fort Collins, CO (United States)

1993-04-01T23:59:59.000Z

134

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

135

Final Manuscript submitted to the Fire Safety Journal Chen et. al., Fire detection using smoke and gas sensors  

E-Print Network (OSTI)

Santa Fe, NM 87505, USA 2 Department of Fire Protection Engineering University of Maryland, College Park Protection Engineering University of Maryland, College Park, MD 20742-3031, USA Abstract Fire detectionFinal Manuscript submitted to the Fire Safety Journal Chen et. al., Fire detection using smoke

136

COMPARATIVE SAFETY ANALYSIS DAT! 16 June 1972  

E-Print Network (OSTI)

thermal battery prematurely". The trees were developed down to events that could not be further developed in the safe arm slide timer and battery timer designs. The analysis shows the new timers do not, in fact (rachet release) design against the new (direct drive) design for both safe arm timer and battery timer

Rathbun, Julie A.

137

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

138

Safety analysis report for packaging (onsite) multicanister overpack cask  

SciTech Connect

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

Edwards, W.S.

1997-07-14T23:59:59.000Z

139

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

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

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

140

SCALE 6: Comprehensive Nuclear Safety Analysis Code System  

SciTech Connect

Version 6 of the Standardized Computer Analyses for Licensing Evaluation (SCALE) computer software system developed at Oak Ridge National Laboratory, released in February 2009, contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package, which is used worldwide. This paper highlights the capabilities of the SCALE system, including continuous-energy flux calculations for processing multigroup problem-dependent cross sections, ENDF/B-VII continuous-energy and multigroup nuclear cross-section data, continuous-energy Monte Carlo criticality safety calculations, Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques, one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations, two- and three-dimensional lattice physics depletion analyses, fast and accurate source terms and decay heat calculations, automated burnup credit analyses with loading curve search, and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

Bowman, Stephen M [ORNL

2011-01-01T23:59:59.000Z

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

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.

Not Available

1992-12-01T23:59:59.000Z

142

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network (OSTI)

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . 15 3.0 REGULATORY IMPACT REVIEW: ECONOMIC AND SOCIOECONOMIC IMPACTS OF THE ALTERNATIVES FOR AMENDMENT 47 TO THE FISHERY MANAGEMENT PLAN FOR THE GROUNDFISH FISHERY OF THE BERING SEA AND ALEUTIAN

143

ENVIRONMENTAL ASSESSMENT / REGULATORY IMPACT REVIEW / FINAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network (OSTI)

ENVIRONMENTAL ASSESSMENT / REGULATORY IMPACT REVIEW / FINAL REGULATORY FLEXIBILITY ANALYSIS Management Plan for Groundfish of the Bering Sea and Aleutian Islands Management Area October 2007 Prepared that bycatch may be exacerbated by the current regulatory closure regulations, as much higher salmon bycatch

144

Deconvolution of variability and uncertainty in the Cassini safety analysis  

SciTech Connect

The standard method for propagation of uncertainty in a risk analysis requires rerunning the risk calculation numerous times with model parameters chosen from their uncertainty distributions. This was not practical for the Cassini nuclear safety analysis, due to the computationally intense nature of the risk calculation. A less computationally intense procedure was developed which requires only two calculations for each accident case. The first of these is the standard 'best-estimate' calculation. In the second calculation, variables and parameters change simultaneously. The mathematical technique of deconvolution is then used to separate out an uncertainty multiplier distribution, which can be used to calculate distribution functions at various levels of confidence.

Kampas, Frank J.; Loughin, Stephen [Lockheed Martin Missiles and Space, P.O. Box 8555, Philadelphia, Pennsylvania 19101 (United States); WAM Systems, 650 Loraine Street, Ardmore, Pennsylvania 19003 (United States)

1998-01-15T23:59:59.000Z

145

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

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

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

146

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

SciTech Connect

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

147

Safety Analysis of Requirements for a Product Family Robyn R. Lutz  

E-Print Network (OSTI)

Safety Analysis of Requirements for a Product Family Robyn R. Lutz Iowa State University and Jet, destatez@collins.rockwell.com Abstract A safety analysis was performed on the software re- quirements for a family of ight instrumentation dis- plays of commercial aircraft. First, an existing Safety Checklist

Lutz, Robyn R.

148

Safety Analysis of Requirements for a Product Family Robyn R. Lutz \\Lambda  

E-Print Network (OSTI)

Safety Analysis of Requirements for a Product Family Robyn R. Lutz \\Lambda Iowa State University and Communication srtockey, destatez@collins.rockwell.com Abstract A safety analysis was performed on the software, an existing Safety Checklist was extended to apply to four­variable models and used to analyze

Lutz, Robyn R.

149

Safety analysis of software product lines using state-based modeling q , Josh Dehlinger a  

E-Print Network (OSTI)

Safety analysis of software product lines using state-based modeling q Jing Liu a , Josh Dehlinger of managing variations and their potential interactions across an entire product line currently hinders safety analysis in safety-critical, software product lines. The work described here contributes to a solution

Lutz, Robyn R.

150

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

151

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

152

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

153

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

154

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

SciTech Connect

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

Not Available

1994-09-01T23:59:59.000Z

155

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.

156

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......errors, computational models (software), management, communication, safety culture, plant ageing, maintenance...Energy Power Plants Probability Radiation Monitoring Radiation Protection Radioactive Hazard Release Safety Management...

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

157

DOE's Approach to Nuclear Facility Safety Analysis and Management  

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

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

158

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

159

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

SciTech Connect

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

160

New enhancements to SCALE for criticality safety analysis  

SciTech Connect

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.


161

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

SciTech Connect

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

162

Deconvolution of variability and uncertainty in the Cassini safety analysis  

SciTech Connect

The standard method for propagation of uncertainty in a risk analysis requires rerunning the risk calculation numerous times with model parameters chosen from their uncertainty distributions. This was not practical for the Cassini nuclear safety analysis, due to the computationally intense nature of the risk calculation. A less computationally intense procedure was developed which requires only two calculations for each accident case. The first of these is the standard {open_quotes}best-estimate{close_quotes} calculation. In the second calculation, variables and parameters change simultaneously. The mathematical technique of deconvolution is then used to separate out an uncertainty multiplier distribution, which can be used to calculate distribution functions at various levels of confidence. {copyright} {ital 1998 American Institute of Physics.}

Kampas, F.J. [Lockheed Martin Missiles and Space, P.O. Box 8555, Philadelphia, Pennsylvania 19101 (United States); Loughin, S. [WAM Systems, 650 Loraine Street, Ardmore, Pennsylvania 19003 (United States)

1998-01-01T23:59:59.000Z

163

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

164

Software Approach to Hazard Detection Using On-line Analysis of Safety Constraints  

E-Print Network (OSTI)

Software Approach to Hazard Detection Using On-line Analysis of Safety Constraints Beth Schroedey. The research here addresses the problem of enhancing software safety through hazard detection. The premise.gatech.edu Abstract Hazard situations in safety-critical systems are typically complex, so there is a need for means

Plale, Beth

165

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

SciTech Connect

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

166

A Conceptual Framework for Semantic Case-based Safety Analysis Olawande Daramola, Tor Stlhane  

E-Print Network (OSTI)

.biffl}@tuwien.ac.at Abstract Hazard and Operability (HAZOP) Analysis and Fail- ure Mode and Effect Analysis (FMEA) are among-based framework for safety analy- sis, which facilitates the reuse of previous HAZOP and FMEA experiences in order application. Key words: Safety analysis, HAZOP, FMEA, ontology, requirements, case-based reasoning, natural

167

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

168

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

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

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

169

Object Oriented Safety Analysis of an Extra High Voltage Substation Bay  

Science Journals Connector (OSTI)

Experiences of application of the object oriented approach to safety analysis of an extra high voltage substation bay are presented. As the first step...

Bartosz Nowicki; Janusz Grski

1998-01-01T23:59:59.000Z

170

Automation of System Safety Analysis: Possibilities and Pitfalls Andrew Galloway, University of York, Heslington, York YO10 5DD UK  

E-Print Network (OSTI)

evolved to support safety analysis work (in this paper, we use the term "safety analysis" to encompass all, safety engineers may only achieve closure at the end of a system's working life, when it is possibleAutomation of System Safety Analysis: Possibilities and Pitfalls Andrew Galloway, University

Pumfrey, David

171

Hazard screening application guide. Safety Analysis Report Update Program  

SciTech Connect

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

none,

1992-06-01T23:59:59.000Z

172

Fault tree synthesis for software design analysis of PLC based safety-critical systems  

SciTech Connect

As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

Koo, S. R.; Cho, C. H. [Corporate R and D Inst., Doosan Heavy Industries and Construction Co., Ltd., 39-3, Seongbok-Dong, Yongin-Si, Gyeonggi-Do 449-795 (Korea, Republic of); Seong, P. H. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Inst. of Science and Technology, 373-3 Guseong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2006-07-01T23:59:59.000Z

173

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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

174

OF-FMEA: an approach to safety analysis of object-oriented software intensive systems  

Science Journals Connector (OSTI)

The paper presents an extension to the common FMEA method in such a way that it can be applied to safety analysis of systems (hardware and software) that are developed using a recently popular object oriented approach. The method makes use of the object ... Keywords: FMEA, formal analysis, safety critical systems

Tadeusz Cichocki; Janusz Grski

2003-01-01T23:59:59.000Z

175

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

176

Comparative analysis of LMFBR licensing in the United States and other countries - notably France. Final report  

SciTech Connect

This report presents an analysis of LMFBR licensing in the United States. It approaches this question broadly, examining first the system in the United States with the various sectors of the nuclear power economy, and the experience of that system in LWR licensing. It then examines the nature of LMFBR safety licensing questions - to the degree that they differ from those of LWR's - and surveys the experience of the United States and other countries in LMFBR safety licensing. Special attention is devoted to the case of France because of the technical leadership which the French program has provided, and because of the apparent efficiency with which French licensing is performed. The French licensing system and LWR licensing experience are examined, and conclusions drawn regarding the reasons for their effectiveness. Finally, a general comparison of the United States and foreign licensing systems is performed, proposals offered during the recent past for changes in the United States system are examined, and possibilities for future changes are suggested.

Golay, M.W.; Castillo, M.

1981-09-30T23:59:59.000Z

177

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

178

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

179

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

180

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

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181

Safety evaluation report related to Babcock and Wilcox Owners Group Plant Reassessment Program: (Final report)  

SciTech Connect

After the accident of Three Mile Island, Unit 2, nuclear power plant owners made a number of improvements to their nuclear facilities. Despite these improvements, the US Nuclear Regulatory Commission (NRC) staff is concerned that the number and complexity of events at Babcock and Wilcox (B and W) nuclear plants have not decreased as expected. This concern was reinforced by the June 9, 1985 total-loss-of-feedwater event at Davis-Besse Nuclear Power Station and the December 26, 1985 overcooling transient at Rancho Seco Nuclear Generating Station. By letter dated January 24, 1986, the Executive Director for Operations (EDO) informed the Chairman of the B and W Owners Group (BWOG) that a number of recent events at B and W-designed reactors have led the NRC staff to conclude that the basic requirements for B and W reactors need to be reexamined. In its February 13, 1986 response to the EDO's letter, the BWOG committed to lead an effort to define concerns relative to reducing the frequency of reactor trips and the complexity of post-trip response in B and W plants. The BWOG submitted a description of the B and W program entitled ''Safety and Performance Improvement Program'' (BAW-1919) on May 15, 1986. Five revisions to BAW-1919 have also been submitted. The NRC staff has reviewed BAW-1919 and its revisions and presents its evaluation in this report. 2 figs., 34 tabs.

Not Available

1987-11-01T23:59:59.000Z

182

Criticality safety analysis of a borated-concrete absorber  

SciTech Connect

Fuel cycle facilities use slab tanks to store fissile solutions, because these tanks have both a high volume-to-floorspace efficiency and an easily verifiable, criticality control (thickness). The results of preliminary criticality analyses using a validated computer code and cross-section library, indicate that a slab tank designed without a solid neutron absorber is not economical in view of process requirements (inventory) and space limitations (layout). A subsequent calculational study assessed the possible increase in the thickness of a single, isolated slab tank using a solid neutron absorber. Finally, an analysis was performed to evaluate the maximum slab thickness for an array of tank/absorbers. The result of these studies showed the potential for expansion of slab tank thickness. 7 refs., 5 figs., 7 tabs.

Funabashi, H.; Oka, T.; Matsumoto, T.; Smolen, G.R. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan); Oak Ridge National Lab., TN (USA))

1989-01-01T23:59:59.000Z

183

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

184

Organic tanks safety program waste aging studies. Final report, Revision 1  

SciTech Connect

Uranium and plutonium production at the Hanford Site produced large quantities of radioactive byproducts and contaminated process chemicals that are stored in underground tanks awaiting treatment and disposal. Having been made strongly alkaline and then subjected to successive water evaporation campaigns to increase storage capacity, the wastes now exist in the physical forms of saltcakes, metal oxide sludges, and aqueous brine solutions. Tanks that contain organic process chemicals mixed with nitrate/nitrite salt wastes might be at risk for fuel-nitrate combustion accidents. This project started in fiscal year 1993 to provide information on the chemical fate of stored organic wastes. While historical records had identified the organic compounds originally purchased and potentially present in wastes, aging experiments were needed to identify the probable degradation products and evaluate the current hazard. The determination of the rates and pathways of degradation have facilitated prediction of how the hazard changes with time and altered storage conditions. Also, the work with aged simulated waste contributed to the development of analytical methods for characterizing actual wastes. Finally, the results for simulants provide a baseline for comparing and interpreting tank characterization data.

Camaioni, D.M.; Samuels, W.D.; Linehan, J.C. [and others

1998-09-01T23:59:59.000Z

185

Multi-Experts Analytic Hierarchy Process for the Sensitivity Analysis of Passive Safety Systems  

E-Print Network (OSTI)

Multi-Experts Analytic Hierarchy Process for the Sensitivity Analysis of Passive Safety Systems YU systems to increase their safety and reliability. However, during accidental scenarios, uncertainties Residual Heat Removal system (RHRs) of the High Temperature Reactor-Pebble Modular (HTR-PM). Key words

Paris-Sud XI, Université de

186

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

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

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

187

Aspects of environmental and safety analysis of fusion reactors  

E-Print Network (OSTI)

This report summarizes the progress made between October 1976 and September 1977 in studies of some environmental and safety considerations in fusion reactor plants. A methodology to assess the admissible occurrence rate ...

Kazimi, Mujid S.

1977-01-01T23:59:59.000Z

188

Hydrogen Safety Project chemical analysis support task: Window C'' volatile organic analysis  

SciTech Connect

This data package contains the results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization of samples for the 101-SY Hydrogen Safety Project. The samples were submitted for analysis by Westinghouse Hanford Company (WHC) under the Technical Project Plan (TPP) 17667 and the Quality Assurance Plan MCS-027. They came from a core taken during Window C'' after the May 1991 gas release event. The analytical procedures required for analysis were defined in the Test Instructions (TI) prepared by the PNL 101-SY Analytical Chemistry Laboratory (ACL) Project Management Office in accordance with the TPP and the QA Plan. The requested analysis for these samples was volatile organic analysis. The quality control (QC) requirements for each sample are defined in the Test Instructions for each sample. The QC requirements outlined in the procedures and requested in the WHC statement of work were followed.

Gillespie, B.M.; Stromatt, R.W.; Ross, G.A.; Hoope, E.A.

1992-01-01T23:59:59.000Z

189

Hydrogen Safety Project chemical analysis support task: Window ``C`` volatile organic analysis  

SciTech Connect

This data package contains the results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization of samples for the 101-SY Hydrogen Safety Project. The samples were submitted for analysis by Westinghouse Hanford Company (WHC) under the Technical Project Plan (TPP) 17667 and the Quality Assurance Plan MCS-027. They came from a core taken during Window ``C`` after the May 1991 gas release event. The analytical procedures required for analysis were defined in the Test Instructions (TI) prepared by the PNL 101-SY Analytical Chemistry Laboratory (ACL) Project Management Office in accordance with the TPP and the QA Plan. The requested analysis for these samples was volatile organic analysis. The quality control (QC) requirements for each sample are defined in the Test Instructions for each sample. The QC requirements outlined in the procedures and requested in the WHC statement of work were followed.

Gillespie, B.M.; Stromatt, R.W.; Ross, G.A.; Hoope, E.A.

1992-01-01T23:59:59.000Z

190

SAFETY ANALYSIS AND INTEGRATION FOR ROBOTIC SYSTEMS -APPLICATION TO A  

E-Print Network (OSTI)

Analysis (FMEA) and Fault Tree Analysis (FTA) which identify potential unit errors resulting in hazards

Guiochet, Jérémie

191

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

192

Use of Fault Tree Analysis for Automotive Reliability and Safety Analysis  

SciTech Connect

Fault tree analysis (FTA) evolved from the aerospace industry in the 1960's. A fault tree is deductive logic model that is generated with a top undesired event in mind. FTA answers the question, ''how can something occur?'' as opposed to failure modes and effects analysis (FMEA) that is inductive and answers the question, ''what if?'' FTA is used in risk, reliability and safety assessments. FTA is currently being used by several industries such as nuclear power and chemical processing. Typically the automotive industries uses failure modes and effects analysis (FMEA) such as design FMEAs and process FMEAs. The use of FTA has spread to the automotive industry. This paper discusses the use of FTA for automotive applications. With the addition automotive electronics for various applications in systems such as engine/power control, cruise control and braking/traction, FTA is well suited to address failure modes within these systems. FTA can determine the importance of these failure modes from various perspectives such as cost, reliability and safety. A fault tree analysis of a car starting system is presented as an example.

Lambert, H

2003-09-24T23:59:59.000Z

193

CORCON-MOD1 preliminary evaluation and application to safety analysis of a large LMFBR plant  

SciTech Connect

The CORCON-MOD1 core material-concrete interaction code, developed at the Sandia Laboratories for LWR safety analysis, was adapted for analyzing a postulated LMFBR core melt accident.

Chen, K.H.; Ray, K.S.

1981-06-30T23:59:59.000Z

194

Fuzzy Failure Rate for Nuclear Power Plant Probabilistic Safety Assessment by Fault Tree Analysis  

Science Journals Connector (OSTI)

Reliability data is essential for a nuclear power plant probabilistic safety assessment by fault tree analysis ... a failure possibility-based reliability algorithm to assess nuclear event reliability data from f...

Julwan Hendry Purba; Jie Lu; Guangquan Zhang

2012-01-01T23:59:59.000Z

195

Development of a safety analysis system for the offshore personnel and equipment transfer process  

E-Print Network (OSTI)

and Effect Analysis (FMEA) was performed. With the FMEA the question "What if?" is asked for each component. For example, "What if the lifting cable fails' ?". Each component was evaluated for failure mode, failure effects on other components... swell or waves stood out as primary factors in the safety of the transfer process. Failure Mode and Effect Anal sis With component failure a recognized factor in the safety of the transfer process, more in-depth analysis was merited. The FMEA...

McKenna, Michael George

2012-06-07T23:59:59.000Z

196

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

Directives, Delegations, and Requirements

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

1997-12-12T23:59:59.000Z

197

Formal Safety analysis of a radiobased railroad crossing using Deductive CauseConsequence  

E-Print Network (OSTI)

#ects analysis (FMEA) and fault tree analysis (FTA). We apply the method to a real world case study: a radio (DCCA). This technique is a formal generalization of well­known safety analysis methods like FMEA [10 by analyzed) than traditional FMEA. We show, that the results of DCCA have the same semantics as those

Reif, Wolfgang

198

Final Meeting Summary ...  

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

Health, Safety, and Environmental Protection Committee November 8, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE November...

199

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

200

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

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

A semiotic analysis of biotechnology and food safety photographs  

E-Print Network (OSTI)

This study evaluated photographs used in Time, Newsweek, and U.S. News and World Report in stories about biotechnology and food safety issues from the years 2000 and 2001. This study implemented a semiotic methodology to determine if the messages...

Norwood, Jennifer Lynn

2006-04-12T23:59:59.000Z

202

Oak Ridge National Laboratory site data for safety-analysis report  

SciTech Connect

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

Fitzpatrick, F.C.

1982-12-01T23:59:59.000Z

203

Safety, Dependability and Performance Analysis of Extended AADL Models  

Science Journals Connector (OSTI)

......analysis Failure mode and effects analysis (FMEA) and fault tree analysis (FTA), model...Wesupporttwopopularhazardanalysistechniques:FMEA and FTA. Both techniques are realized...symbolic model checking [33, 34]. (i) FMEA is an inductive technique that starts by......

Marco Bozzano; Alessandro Cimatti; Joost-Pieter Katoen; Viet Yen Nguyen; Thomas Noll; Marco Roveri

2011-05-01T23:59:59.000Z

204

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

205

Safety Analysis of an Airbag System using Probabilistic FMEA and Probabilistic Counter Examples  

E-Print Network (OSTI)

Safety Analysis of an Airbag System using Probabilistic FMEA and Probabilistic Counter Examples Failure mode and effects analysis (FMEA) is a technique to reason about possible system hazards that result from system or system component failures. Traditionally, FMEA does not take the probabilities

Leue, Stefan

206

Safety Analysis of an Airbag System using Probabilistic FMEA and Probabilistic Counterexamples  

E-Print Network (OSTI)

Safety Analysis of an Airbag System using Probabilistic FMEA and Probabilistic Counterexamples H analysis (FMEA) is a technique to reason about possible system hazards that result from system or system component failures. Tradition- ally, FMEA does not take the probabilities with which these failures may

Leue, Stefan

207

Overheads, Safety Analysis and Engineering FY 1995 Site Support Program Plan WBS 6.3.5  

SciTech Connect

The Safety Analysis & Engineering (SA&E) department provides core competency for safety analysis and risk documentation that supports achievement of the goals and mission as described in the Hanford Mission Plan, Volume I, Site Guidance (DOE-RL 1993). SA&E operations are integrated into the programs that plan and conduct safe waste management, environmental restoration, and operational activities. SA&E personnel are key members of task teams assigned to eliminate urgent risks and inherent threats that exist at the Hanford Site. Key to ensuring protection of public health and safety, and that of onsite workers, are the products and services provided by the department. SA&E will continue to provide a leadership role throughout the DOE complex with innovative, cost-effective approaches to ensuring safety during environmental cleanup operations. The SA&E mission is to provide support to direct program operations through safety analysis and risk documentation and to maintain an infrastructure responsive to the evolutionary climate at the Hanford Site. SA&E will maintain the appropriate skills mix necessary to fulfill the customers need to conduct all operations in a safe and cost-effective manner while ensuring the safety of the public and the onsite worker.

DiVincenzo, E.P.

1994-09-27T23:59:59.000Z

208

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

209

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

SciTech Connect

The ability to better characterize and quantify safety margins is important to improved decision making about nuclear power plant design, operation, and plant life extension. As research and development (R&D) in the light-water reactor (LWR) Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a station blackout wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario.

Curtis Smith; Diego Mandelli; Cristian Rabiti

2013-11-01T23:59:59.000Z

210

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

SciTech Connect

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

211

FAQS Gap Analysis Qualification Card Nuclear Safety Specialist  

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

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

212

Safety Analysis (SA) of the decontamination facility, Building 419, at the Lawrence Livermore National Laboratory  

SciTech Connect

This safety analysis was performed for the Manager, Plant Services at LLNL and fulfills the requirements of DOE Order 5481.1. The analysis was based on field inspections, document review, computer calculations, and extensive input from Waste Management personnel. It was concluded that the maximum quantities of radioactive materials that safety procedures allow to be handled in this building do not pose undue risks on- or off-site even in postulated severe accidents. Risk from the various hazards at this facility vary from low to moderate as specified in DOE Order 5481.1. Recommendations are made for improvements that will reduce risks even further.

Odell, B.N.

1980-06-17T23:59:59.000Z

213

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

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

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

214

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.

215

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500MWe PFBR  

Science Journals Connector (OSTI)

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermalhydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.

T. Sajith Mathews; M. Ramakrishnan; U. Parthasarathy; A. John Arul; C. Senthil Kumar

2008-01-01T23:59:59.000Z

216

Vertically integrated analysis of human DNA. Final technical report  

SciTech Connect

This project has been oriented toward improving the vertical integration of the sequential steps associated with the large-scale analysis of human DNA. The central focus has been on an approach to the preparation of {open_quotes}sequence-ready{close_quotes} maps, which is referred to as multiple-complete-digest (MCD) mapping, primarily directed at cosmid clones. MCD mapping relies on simple experimental steps, supported by advanced image-analysis and map-assembly software, to produce extremely accurate restriction-site and clone-overlap maps. We believe that MCD mapping is one of the few high-resolution mapping systems that has the potential for high-level automation. Successful automation of this process would be a landmark event in genome analysis. Once other higher organisms, paving the way for cost-effective sequencing of these genomes. Critically, MCD mapping has the potential to provide built-in quality control for sequencing accuracy and to make possible a highly integrated end product even if there are large numbers of discontinuities in the actual sequence.

Olson, M.

1997-10-01T23:59:59.000Z

217

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

SciTech Connect

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

218

NASA/TM-2007-214856 Safety and Performance Analysis of the  

E-Print Network (OSTI)

March 2007 NASA/TM-2007-214856 Safety and Performance Analysis of the Non-Radar Oceanic National Institute of Aerospace, Hampton, Virginia #12;The NASA STI Program Office . . . in Profile Since its founding, NASA has been dedicated to the advancement of aeronautics and space science. The NASA

Muñoz, César A.

219

NASA/TM-2009-215768 A Mathematical Basis for the Safety Analysis of  

E-Print Network (OSTI)

June 2009 NASA/TM-2009-215768 A Mathematical Basis for the Safety Analysis of Conflict Prevention, Virginia Gilles Dowek Ecole Polytechnique, France #12;NASA STI Program . . . in Profile Since its founding, NASA has been dedicated to the advancement of aeronautics and space science. The NASA scientific

Maddalon, Jeffrey M.

220

Combining Functional and Structural Reasoning for Safety Analysis of Electrical Designs  

E-Print Network (OSTI)

in detail. FLAME has been developed over several years, and is capable of composing an FMEA report for many Failure mode effects analysis (FMEA) of a design involves the investigation and assessment of the effects, electronic and mechanical systems are being combined in safety-critical applications. Automation of FMEA

Snooke, Neal

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

Results from One- and Two- Phase Fluid Flow Calculations within the Preliminary Safety Analysis of the Gorleben Site - 13310  

SciTech Connect

Rock salt is one of the possible host rock formations for the disposal of high-level radioactive wastes in Germany. The Preliminary Safety Analysis of the Gorleben Site (Vorlaeufige Sicherheitsanalyse Gorleben, VSG) evaluates the long-term safety of a hypothetical repository in the salt dome of Gorleben, Germany. A mature repository concept and detailed knowledge of the site allowed a detailed process analysis within the project by numerical modeling of single-phase and two-phase flow. The possibility of liquid transport from the shafts to the emplacement drifts is one objective of the present study. Also, the implications of brine inflow on radionuclide transport and gas generation are investigated. Pressure build-up due to rock convergence and gas generation, release of volatile radionuclides from the waste and pressure-driven contaminant transport were considered, too. The study confirms that the compaction behavior of salt grit backfill is one of the most relevant factors for the hydrodynamic evolution of the repository and the transport of contaminants. Due to the interaction between compaction, saturation and pore pressure, complex flow patterns evolve. Emplacement drifts serve as gas sinks or sources at different times. In most calculation cases, the backfill reaches its final porosity after a few hundred years. The repository is then sealed and radionuclides can only be transported by diffusion in the liquid phase. Estimates for the final porosity of compacted backfill range between 0 % and 2 %. The exact properties of the backfill regarding single- and two-phase flow are not well known for this porosity range. The study highlights that this uncertainty has a profound impact on flow and transport processes over long time-scales. Therefore, more research is needed to characterize the properties of crushed salt grit at low porosities or to reduce the adverse effects of possible higher porosities by repository optimization. (authors)

Kock, Ingo; Larue, Juergen; Fischer, Heidi; Frieling, Gerd; Navarro, Martin; Seher, Holger [Department of Final Disposal, GRS mbH, Schwertnergasse 1, 50667 Cologne (Germany)] [Department of Final Disposal, GRS mbH, Schwertnergasse 1, 50667 Cologne (Germany)

2013-07-01T23:59:59.000Z

222

Plastic heliostat and heliostat enclosure analysis. Final report  

SciTech Connect

The conceptual design and cost analysis report of an enclosed plastic heliostat for a 50-MW/sub e/ central receiver solar thermal electric power plant are presented. The purpose of the study was to analyze the most recent design of the Boeing enclosed plastic heliostat for cost and compare results with a reference second generation glass heliostat case provided by Sandia National Laboratories, Livermore (SNLL). In addition, sensitivities of busbar energy costs to variations in capital cost (installed cost), operation and maintenance cost and overall reflectivity were evaluated.

Berry, M.J.

1984-12-01T23:59:59.000Z

223

Aerosol analysis for the regional air pollution study. Final report  

SciTech Connect

The design and operation of an aerosol sampling and analysis program implemented during the 1975 to 1977 St. Louis Regional Air Pollution Study is described. A network of ten samplers were operated at selected sites in the St. Louis area and the total mass and elemental composition of the collected particulates were determined. Sampling periods of 2 to 24 hours were employed. The samplers were capable of collecting aerosol particles in two distinct size ranges corresponding to fine (< 2.4 ..mu..m diameter) and coarse (> 2.4 ..mu..m diameter) particles. This unique feature allowed the separation of the particulate samples into two distinct fractions with differing chemical origins and health effects. The analysis methods were also newly developed for use in the St. Louis RAPS study. Total particulate mass was measured by a beta-particle attenuation method in which a precision of +- 5 ..mu..m/cm/sup 2/ could be obtained in a one minute measurement time. Elemental compositions of the samples were determined using an energy dispersive x-ray fluorescence method in which detectable limits of 5 ng/cm/sup 2/ or less were routinely achieved for elements ranging in atomic number from Al to Pb. The advantages of these analytical methods over more conventional techniques arise from the ability to automate the measurements. During the course of the two year study, a total of more than 35,000 individual samples were processed and a total of 28 concentrations measured for each sample.

Jaklevic, J.M.; Gatti, R.C.; Goulding, F.S.; Loo, B.W.; Thompson, A.C.

1980-05-01T23:59:59.000Z

224

Final Environmental Assessment/ Regulatory Impact Review/ Initial Regulatory Flexibility Analysis for Amendment 93 to the Fishery  

E-Print Network (OSTI)

Final Environmental Assessment/ Regulatory Impact Review/ Initial Regulatory Flexibility Analysis Review/Initial Regulatory Flexibility Analysis analyzing proposed management measures that would apply for Amendment 93 to the Fishery Management Plan for Groundfish of the Gulf of Alaska Chinook Salmon Prohibited

225

Caucasus Seismic Information Network: Data and Analysis Final Report  

SciTech Connect

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

226

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

227

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

SciTech Connect

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

228

Waste Tank Organic Safety Project: Analysis of liquid samples from Hanford waste tank 241-C-103  

SciTech Connect

A suite of physical and chemical analyses has been performed in support of activities directed toward the resolution of an Unreviewed Safety Question concerning the potential for a floating organic layer in Hanford waste tank 241-C-103 to sustain a pool fire. The analysis program was the result of a Data Quality Objectives exercise conducted jointly with staff from Westinghouse Hanford Company and Pacific Northwest Laboratory (PNL). The organic layer has been analyzed for flash point, organic composition including volatile organics, inorganic anions and cations, radionuclides, and other physical and chemical parameters needed for a safety assessment leading to the resolution of the Unreviewed Safety Question. The aqueous layer underlying the floating organic material was also analyzed for inorganic, organic, and radionuclide composition, as well as other physical and chemical properties. This work was conducted to PNL Quality Assurance impact level III standards (Good Laboratory Practices).

Pool, K.H.; Bean, R.M.

1994-03-01T23:59:59.000Z

229

A case study on effectiveness of structural reliability analysis in nuclear reactor safety assessment  

Science Journals Connector (OSTI)

Problems on reliability of structural integrity occupy an important position in various aspects of nuclear reactor safety. In the present paper, an effective method for quantitative evaluation of structural reliability based on stress strength model is developed with the objectives of taking a larger number of factors into the evaluation than before and giving useful results within moderate computing time. The method is applied to the reliability analysis of PWR pressure vessels. The results show the relative importance of inspection as well as the parameter uncertainty for assuring the reliability of the structure, although analysis is limited within the scope of linear elastic fracture mechanics (LEFM). This case study also shows that the analysis of structural reliability is effective for safety assessment of nuclear power plants in general and possibly for the improvements of the consistency in the design code.

A. Yamaguchi; S. Kondo; Y. Togo

1983-01-01T23:59:59.000Z

230

Subsistence restoration project: Food safety testing. Exxon Valdez Oil Spill Restoration Project. Final report restoration project 94279  

SciTech Connect

The goal of this project was to restore the confidence of subsistence users in their abilities to determine the safety of their resources. Methods included community meetings, collection and testing of subsistence resources samples for hydrocarbon contamination, accompanying community representatives on tours of the laboratory where tests were conducted and informational newsletters. Over the two years of the project combined, 228 composite samples of edible tissues from shellfish were tested. The bile of forty rockfish, six sockeye salmon, twelve seals, twenty-three ducks were tested for the presence of hydrocarbon metabolites. Edible tissue (blubber) from seals was also tested. Generally, the tests showed such low levels of hydrocarbons and their metabolites, as to be within the test`s margin of error. The project was partly successful in disseminating the subsistence food safety advice of the Oil Spill Health Task Force and in improving the level of trust in the results of hydrocarbon tests on the resources.

Miraglia, R.A.; Chartrand, A.W.

1997-05-01T23:59:59.000Z

231

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

SciTech Connect

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

Herborn, D.I.

1993-11-01T23:59:59.000Z

232

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

233

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

SciTech Connect

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

Boyd, G.J.; Lewis, S.R.; Summitt, R.L. [Safety and Reliability Optimization Services (SAROS), Inc., Knoxville, TN (United States)

1991-12-31T23:59:59.000Z

234

Laser Safety and Hazardous Analysis for the ARES (Big Sky) Laser System  

SciTech Connect

A laser safety and hazard analysis was performed for the ARES laser system based on the 2000 version of the American National Standards Institute's (ANSI) Standard Z136.1,for Safe Use of Lasers and the 2000 version of the ANSI Standard Z136.6, for Safe Use of Lasers Outdoors. The ARES laser system is a Van/Truck based mobile platform, which is used to perform laser interaction experiments and tests at various national test sites.

AUGUSTONI, ARNOLD L.

2003-01-01T23:59:59.000Z

235

Safety analysis report for packaging, onsite, long-length contaminated equipment transport system  

SciTech Connect

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks.

McCormick, W.A.

1997-05-09T23:59:59.000Z

236

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

237

Final Accepted Version Collision risk-capacity tradeoff analysis of an en-route  

E-Print Network (OSTI)

Final Accepted Version Collision risk-capacity tradeoff analysis of an en-route corridor model YE-route corridor. This paper analyzes the collision risk-capacity tradeoff using combined discrete variables can improve the rate and stability of the corridor with low risks of loss of separation. Keywords

238

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)

Health and Safety Impacts of Nuclear, Geothermal, and Fossil- Fuel3 of HEALTH AND SAFETY IMPACTS OF FOSSIL-FUEL NUCLEAR,HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL

Nero, A.V.

2010-01-01T23:59:59.000Z

239

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)

nuclear tors. for of of These studies can examine safety systems or safety research programsnuclear power plants, and at risk. to reduce population The Light-water Reactor Safety Research Program

Nero, A.V.

2010-01-01T23:59:59.000Z

240

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

SciTech Connect

This document contains the completed safety analysis which establishes the safety envelope for performing the mixer pump process test in Tank 241-AZ-101. This process test is described in TF-210-OTP-001. All equipment necessary for the mixer pump test has been installed by Project W-151. The purpose of this document is to describe and analyze the mixer pump test for Aging Waste Facility (AWF) Tank 241-AZ-101 and to address the 'yes/maybe' responses marked for evaluation questions identified in Unreviewed Safety Question Evaluation (USQE) TF-94-0266. The scope of this document is limited to the performance of the mixer pump test for Tank 241-AZ-101. Unreviewed Safety Question Determination (USQD) TF-96-0018 verified that the installation of two mixer pumps into Tank 241-AZ-101 was within the current Tank Waste Remediation System (TWRS) Authorization Basis. USQDs TF-96-0461, TF-96-0448, and TF-96-0805 verified that the installation of the in-tank video camera, thermocouples, and Ultrasonic Interface Level Analyzer (URSILLA), respectively, were within the current TWRS Authorization Basis. USQD TF-96-1041 verified that the checkout testing of the installed equipment was within the current TWRS Authorization Basis. Installation of the pumps and equipment has been completed. An evaluation of safety considerations associated with operation of the mixer pumps for the mixer pump test is provided in this document. This document augments the existing AWF authorization basis as defined in the Interim Safety Basis (Stahl 1997), and as such, will use the existing Interim Operational Safety Requirements (IOSRs) of Heubach 1996 to adequately control the mixer pump test. The hazard and accident analysis is limited to the scope and impact of the mixer pump test, and therefore does not address hazards already addressed by the current AWF authorization basis. This document does not evaluate removal of the mixer pumps. Safety considerations for removal of the pumps will be addressed by separate safety documentation once that portion of the mission is defined. The mixer pump test has been evaluated to cover the use of either the existing ventilation system (241-A-702) or the ventilation system upgrade provided by Project W-030. Analysis of Project W-030 is outside of the scope of this document and is addressed in HNF-SD-WM-SARR-039 (Draft) which, should the W-030 system be in service at the time of the mixer pump test, will have been approved and made a part of the TWRS authorization basis. The test will use two high-capacity mixer pumps in various configurations and modes to demonstrate solids mobilization of waste in Tank 241-AZ-101. The information and experience gained during the test will provide data for comparison with sludge mobilization prediction models; provide data to estimate the number, location, and cycle times of the mixer pumps; and provide indication of the effects of mixer pump operation on the AWF tank systems and components. The slurry produced will be evaluated for future pretreatment processing. This process test does not transfer waste from the tank; the waste is mixed and confined within the existing system. At the completion of the mixer pump test, the mixer pumps will be stopped and normal tank operations, maintenance, and surveillance will continue. Periodic rotation of the mixer pumps and motor shafts, along with bearing greasing, is required to maintain the pumps following the mixer pump test.

HAMMOND DM; HARRIS JP; MOUETTE P

1997-06-09T23:59:59.000Z

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.


241

Organic Tanks Safety Program: Advanced organic analysis FY 1996 progress report  

SciTech Connect

Major focus during the first part of FY96 was to evaluate using organic functional group concentrations to screen for energetics. Fourier transform infrared and Raman spectroscopy would be useful screening tools for determining C-H and COO- organic content in tank wastes analyzed in a hot cell. These techniques would be used for identifying tanks of potential safety concern that may require further analysis. Samples from Tanks 241-C-106 and -C-204 were analyzed; the major organic in C-106 was B2EHPA and in C-204 was TBP. Analyses of simulated wastes were also performed for the Waste Aging Studies Task; organics formed as a result of degradation were identified, and the original starting components were monitored quantitatively. Sample analysis is not routine and required considerable methods adaptation and optimization. Several techniques have been evaluated for directly analyzing chelator and chelator fragments in tank wastes: matrix-assisted laser desorption/ionization time-of-flight mass spectrometry and liquid chromatography with ultraviolet detection using Cu complexation. Although not directly funded by the Tanks Safety Program, the success of these techniques have implications for both the Flammable Gas and Organic Tanks Safety Programs.

NONE

1996-09-01T23:59:59.000Z

242

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

SciTech Connect

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

243

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

244

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor  

Science Journals Connector (OSTI)

The 500MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and SteamWater System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.10.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ?1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using CRAFT software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by ? factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ?1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

A. John Arul; C. Senthil Kumar; S. Athmalingam; Om Pal Singh; K. Suryaprakasa Rao

2006-01-01T23:59:59.000Z

245

Electrical Safety Occurrences | Department of Energy  

Office of Environmental Management (EM)

Electrical Safety Occurrences Electrical Safety Occurrences June 26, 2014 Monthly Analysis of Electrical Safety Occurrences - April 2013 An analysis of the Occurrence Reporting and...

246

Electrical Safety Occurrences | Department of Energy  

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

Electrical Safety Occurrences Electrical Safety Occurrences September 20, 2011 Monthly Analysis of Electrical Safety Occurrences - August 2011 An analysis of the Occurrence...

247

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.  

SciTech Connect

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

2010-03-01T23:59:59.000Z

248

Energy Engineering Analysis Program, Wuerzburg Military Community. Executive summary. Final report  

SciTech Connect

This final report is submitted in accordance with the Schedule of Title I Services for Contract DACA 90-81-C-0094 Energy Engineering Analysis Program FY 81 OMA, EEAP 007, Aschaffenburg, Wuerzburg, and Schweinfurt Military Communities, and as amended by Addenda Nos. 1, 2, and 3 to Appendix A and the resume of Negotiations. The purpose of the Energy Engineering Analysis Program (EEAP) is to develop a comprehensive plan for the use of energy and to identify energy conservation projects at each of the military communities.

NONE

1984-06-01T23:59:59.000Z

249

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.

250

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

SciTech Connect

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

NONE

1990-09-19T23:59:59.000Z

251

Safety Analysis and Certification of Open Distributed Systems P. M. Conmy; Department of Computer Science, University of York, York, YO10 5DD U.K.  

E-Print Network (OSTI)

Safety Analysis and Certification of Open Distributed Systems P. M. Conmy; Department of Computer to the safety analysis and certification of avionics computer systems. At present aircraft computing systems are to the way avionics computer systems are analysed and certified. At present analysis techniques are based

Nicholson, Mark

252

Health, Safety and Environmental Protection Committee Page 1  

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

Safety and Environmental Protection Committee Page 1 Final Meeting Summary April 19, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION...

253

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

254

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

255

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

256

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

National Nuclear Security Administration (NNSA)

Services, Washington, D.C., 1968. MACCS2 Guidance Report June 2004 Final Report 9-8 Smith, 1995, B.J. Smith and K.K. Taylor, Safety Analysis in Support of the Environmental...

257

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

258

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

SciTech Connect

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

DeHart, M.D.

1999-08-01T23:59:59.000Z

259

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

260

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

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

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

262

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.

263

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

264

The Front Lines of Patient Safety  

E-Print Network (OSTI)

patient safety · Incident Reporting · Root Cause Analysis · FMEA · Culture of Patient Safety Survey

Soloveichik, David

265

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

266

Fuel-Coolant-Interaction modeling and analysis work for the High Flux Isotope Reactor Safety Analysis Report  

SciTech Connect

A brief historical background and a description of short- and long-term task plan development for effective closure of this important safety issue for the HFIR are given. Short-term aspects deal with Fuel-Coolant-Interaction (FCI) issues experimentation, modeling, and analysis for the flow-blockage-induced steam explosion events in direct support of the SAR. Long-term aspects deal with addressing FCI issues resulting from other accidents in conjunction with issues dealing with aluminum ignition, which can result in an order of magnitude increase in overall energetics. Problem formulation, modeling, and computer code simulation for the various phases of steam explosions are described. The evaluation of core melt initiation propagation, and melt superheat are described. Core melt initiation and propagation have been studied using simple conservative models as well as from modeling and analysis using RELAP5. Core debris coolability, heatup, and melting/freezing aspects have been studied by use of the two-dimensional melting/freezing analysis code 2DKO, which was also benchmarked with MELCOR code predictions. Descriptions are provided for the HM, BH, FCIMOD, and CTH computer codes that have been implemented for studying steam explosion energetics from the standpoint of evaluating bounding loads by thermodynamic models or best-estimate loads from one- and two-dimensional simulations of steam explosion energetics. Vessel failure modeling and analysis was conducted using the principles of probabilistic fracture mechanics in conjunction with ADINA code calculations. Top head bolts failure modeling has also been conducted where the failure criterion was based upon stresses in the bolts exceeding the material yield stress for a given time duration. Missile transport modeling and analysis was conducted by setting up a one-dimensional mathematical model that accounts for viscous dissipation, virtual mass effects, and material inertia.

Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Chang, S.J.; Freels, J.; Gat, U.; Lepard, B.L.; Gwaltney, R.C.; Luttrell, C.; Kirkpatrick, J.

1993-07-01T23:59:59.000Z

267

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

268

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

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

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

269

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

SciTech Connect

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

270

Facility Safety  

Directives, Delegations, and Requirements

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

1996-10-24T23:59:59.000Z

271

Facility Safety  

Directives, Delegations, and Requirements

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

1995-11-16T23:59:59.000Z

272

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

SciTech Connect

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

273

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

274

Nuclear Engineer (Criticality Safety)  

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

This position is located in the Nuclear Safety Division (NSD) which has specific responsibility for managing the development, analysis, review, and approval of non-reactor nuclear facility safety...

275

Uncertainty analysis of criticality safety for the plate type fuel assembly storage rack  

Science Journals Connector (OSTI)

To evaluate the criticality safety of the fresh and the spent fuel storage racks in an open pool type research reactor designed by KAERI, the upper subcriticality limit (USL) analysis was carried out. First, the bias and its uncertainty of MCNP code system with ENDF/B-VII library were evaluated using the calculation results of the 183 benchmark experiments. The criticality calculations for the fuel storage rack are carried out under a normal state, an increased water temperature, a fuel assembly drop, and an eccentric insertion which can affect the criticality. Considering biases and uncertainties for the MCNP code system, abnormal conditions, and the manufacturing tolerance of the cell tube thickness, the USL value that can guarantee sufficient subcriticality is determined. It was found that the criticality of the fresh and the spent fuel storage racks currently designed satisfy the USL condition. Additionally, it was concluded that the pitch size of a fresh fuel storage rack can be reduced for efficient space availability, and even under a worst case in which the fresh storage rack is surrounded by a lower water density and the smallest pitch size satisfies the USL conditions.

Tae Young Han; Chang Je Park; Byung Chul Lee; Jae Man Noh

2013-01-01T23:59:59.000Z

276

Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications  

SciTech Connect

The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

2014-07-01T23:59:59.000Z

277

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

278

Advanced organic analysis and analytical methods development: FY 1995 progress report. Waste Tank Organic Safety Program  

SciTech Connect

This report describes the work performed during FY 1995 by Pacific Northwest Laboratory in developing and optimizing analysis techniques for identifying organics present in Hanford waste tanks. The main focus was to provide a means for rapidly obtaining the most useful information concerning the organics present in tank waste, with minimal sample handling and with minimal waste generation. One major focus has been to optimize analytical methods for organic speciation. Select methods, such as atmospheric pressure chemical ionization mass spectrometry and matrix-assisted laser desorption/ionization mass spectrometry, were developed to increase the speciation capabilities, while minimizing sample handling. A capillary electrophoresis method was developed to improve separation capabilities while minimizing additional waste generation. In addition, considerable emphasis has been placed on developing a rapid screening tool, based on Raman and infrared spectroscopy, for determining organic functional group content when complete organic speciation is not required. This capability would allow for a cost-effective means to screen the waste tanks to identify tanks that require more specialized and complete organic speciation to determine tank safety.

Wahl, K.L.; Campbell, J.A.; Clauss, S.A. [and others

1995-09-01T23:59:59.000Z

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

The safety climate of a Department of Energy nuclear facility: A sociotechnical analysis  

SciTech Connect

Government- and public-sponsored groups are increasingly demanding greater accountability by the Department of Energy`s weapons complex. Many of these demands have focused on the development of a positive safety climate, one that not only protects workers onsite, but also the surrounding populace and environment as well. These demands are, in part, a response to findings which demonstrate a close linkage between actual organizational safety performance and the organization`s safety climate, i.e., the collective attitudes employees hold concerning the level of safety in their organization. This paper describes the approach taken in the systematic assessment of the safety climate at EG&G Rocky Flats Plant (RFP).

Johnson, A.E.; Harbour, J.L.

1993-06-01T23:59:59.000Z

282

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

283

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

284

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

285

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

286

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

287

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

SciTech Connect

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

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

2014-06-01T23:59:59.000Z

288

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

SciTech Connect

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

289

Final Meeting Summary Page 1  

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

October 11, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE October 11, 2012 Richland, WA Topics in this Meeting Summary...

290

Final Meeting Summary Page 1  

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

August 9, 2012 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE August 9, 2012 Richland, WA Topics in this Meeting Summary Opening...

291

Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities; Yucca Mountain Site Characterization Project  

SciTech Connect

This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10{sup {minus}11}/yr to 10{sup {minus}5}/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10{sup {minus}9}/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution.

Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States); Laub, T.W. [Sandia National Labs., Albuquerque, NM (United States)

1992-06-01T23:59:59.000Z

292

Safety First Safety Last Safety Always Safety Shoes  

E-Print Network (OSTI)

Safety First Safety Last Safety Always Safety Shoes and Boots Safety Tip #21 Don't let your day guards) can be used in conjunction with standard safety shoes. Safety boots Safety boots come in many varieties, and which you will use will depend on the specific hazards you face. Boots offer more protection

Minnesota, University of

293

RESEARCH SAFETY RADIATION SAFETY  

E-Print Network (OSTI)

and Communications Manager (951) 827-6303 janette.ducut@ucr.edu Beiwei Tu, MS, CIH, CSP Safety and Industrial Hygiene, CSP Laboratory Safety Compliance Specialist (951) 827-2528 sarah.meyer@ucr.edu (vacant) Integrated

294

A Comparison of Two Approaches to Safety Analysis Based on Use Cases  

Science Journals Connector (OSTI)

Engineering has a long tradition in analyzing the safety of mechanical, electrical and electronic systems. Important methods like HazOp and FMEA have also been adopted by the software ... by the software communit...

Tor Stlhane; Guttorm Sindre

2007-01-01T23:59:59.000Z

295

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

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

Energy Institute (NEI) Attachment, Integrated Safety Analysis Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis This paper addresses why the use of an...

296

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

297

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

298

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

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

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

299

Safety First Safety Last Safety Always General site safety  

E-Print Network (OSTI)

Safety First Safety Last Safety Always General site safety During the course of construction barrier at least 5 feet (1.5m) high having a fire-resistance rating of at least one half hour. Site Safety and Clean-up Safety Tip #20 Safety has no quitting time. All contractors should clean up their debris, trash

Minnesota, University of

300

Safety First Safety Last Safety Always Safety Tip #22  

E-Print Network (OSTI)

Safety First Safety Last Safety Always Safety Tip #22 Mowing Operations Mowing unsafely just doesn for out-of-control vehicles. Wear hearing protection and a safety vest. Wear a hard hat and safety goggles of this safety tip sheet. Please refrain from reading the information verbatim--paraphrase it instead

Minnesota, University of

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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to obtain the most current and comprehensive results.


301

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

302

An analysis of firefighter personal safety alarm effectiveness on the fire ground  

Science Journals Connector (OSTI)

For firefighters in the line of duty the last line of defense and chance for rescue oftentimes relies on the effectiveness of their Personal Alert Safety System (PASS) devices. When activated a PASS device emits an alarm signal to notify others that a firefighter is in distress. However there have been notable instances where PASS devices have confused rescue personnel or created a more hazardous situation for instance when noise interference originating from other objects is involved. This research compiles data from various sources for example firefighter near miss reports and National Institute for Occupational Safety and Health (NIOSH) fatality reports regarding PASS device effectiveness. The research will investigate the causes of confusion and danger as well as take a look at the situations where the device achieved its goal and was able to save a life. The implications of discovering how interfering noises can render PASS devices ineffective could save several lives in the future and ultimately lead to increased firefighter safety.

Kyle Ford; Mudeer Habeeb; Joelle Suits; Mustafa Abbasi; Ofodike Ezekoye

2013-01-01T23:59:59.000Z

303

Environment, Safety, and Health Self-Assessment Report, Fiscal Year 2008  

E-Print Network (OSTI)

and the 4th Annual Safety Culture Survey conducted by HealthFinally, results of the Safety Culture Survey indicate thatawareness and promotes safety culture within the Division.

Chernowski, John

2009-01-01T23:59:59.000Z

304

Safety Assurance for ATR Irradiations  

SciTech Connect

The Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL) is the worlds premiere test reactor for performing high fluence, large volume, irradiation test programs. The ATR has many capabilities and a wide variety of tests are performed in this truly one of a kind reactor, including isotope production, simple self-contained static capsule experiments, instrumented/controlled experiments, and loop testing under pressurized water conditions. Along with the five pressurized water loops, ATR may also have gas (temperature controlled) lead experiments, fuel boosted fast flux experiments, and static sealed capsules all in the core at the same time. In addition, any or all of these tests may contain fuel or moderating materials that can affect reactivity levels in the ATR core. Therefore the safety analyses required to ensure safe operation of each experiment as well as the reactor itself are complex. Each test has to be evaluated against stringent reactor control safety criteria, as well as the effects it could have on adjacent tests and the reactor as well as the consequences of those effects. The safety analyses of each experiment are summarized in a document entitled the Experiment Safety Assurance Package (ESAP). The ESAP references and employs the results of the reactor physics, thermal, hydraulic, stress, seismic, vibration, and all other analyses necessary to ensure the experiment can be irradiated safely in the ATR. The requirements for reactivity worth, chemistry compatibilities, pressure limitations, material issues, etc. are all specified in the Technical Safety Requirements and the Upgraded Final Safety Analysis Report (UFSAR) for the ATR. This paper discusses the ESAP process, types of analyses, types of safety requirements and the approvals necessary to ensure an experiment can be safely irradiated in the ATR.

S. Blaine Grover

2006-10-01T23:59:59.000Z

305

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

SciTech Connect

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

306

Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory. TRU curium shipping container  

SciTech Connect

An analytical evaluation of the Oak Ridge National Laboratory Transuranium (TRU) Curium Shipping Container was made to demonstrate its compliance with the regulations governing offsite shipment of packages containing radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations.

Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Aramayo, G.A.

1980-06-01T23:59:59.000Z

307

Component Failure Behaviour: Patterns and Reuse in Automated System Safety Analysis  

E-Print Network (OSTI)

using a pool of shared information and energy resources. In practice, one of the major challenges together a number of currently standalone functions in a common platform where functions are implemented towards implementation of distributed active safety in vehicles is automatic control of the steering

Boyer, Edmond

308

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

SciTech Connect

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

309

E-Print Network 3.0 - analysis process final Sample Search Results  

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

Series Analysis: Univariate and Multivariate Methods (2nd edition, 2006) by William W.S. Wei... series parameters, ARMA models, trend analysis, model identification,...

310

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

311

Facility Safety  

Directives, Delegations, and Requirements

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

2013-06-21T23:59:59.000Z

312

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

313

Expectations on Documented Safety Analysis for Deactivated Inactive Nuclear Facilities in a State of Long Term Surveillance & Maintenance or Decommissioning  

SciTech Connect

DOE promulgated 10 CFR 830 ''Nuclear Safety Management'' on October 10, 2000. Section 204 of the Rule requires that contractors at DOE hazard category 1, 2, and 3 nuclear facilities develop a ''Documented Safety Analysis'' (DSA) that summarizes the work to be performed, the associated hazards, and hazard controls necessary to protect workers, the public, and the environment. Table 2 of Appendix A to the rule has been provided to ensure that DSAs are prepared in accordance with one of the available predetermined ''safe harbor'' approaches. The table presents various acceptable safe harbor DSAs for different nuclear facility operations ranging from nuclear reactors to decommissioning activities. The safe harbor permitted for decommissioning of a nuclear facility encompasses methods described in DOE-STD-1 120-98, ''Integration of Environment, Safety and Health into Facility Disposition Activities,'' and provisions in 29 CFR 1910.120 or 29 CFR 1926.65 (HAZWOPER). Additionally, an evaluation of public safety impacts and development of necessary controls is required when the facility being decommissioned contains radiological inventory or contamination exceeding the Rule's definition for low-level residual fixed radioactivity. This document discusses a cost-effective DSA approach that is based on the concepts of DOE-STD-I 120 and meets the 10 CFR 830 safe harbor requirements for both transition surveillance and maintenance as well as decommissioning. This DSA approach provides continuity for inactive Hanford nuclear facilities that will eventually transition into decommissioning. It also uses a graded approach that meets the expectations of DOE-STD-3011 and addresses HAZWOPER requirements to provide a sound basis for worker protection, particularly where intrusive work is being conducted.

JACKSON, M.W.

2002-05-01T23:59:59.000Z

314

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

SciTech Connect

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

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

1980-01-01T23:59:59.000Z

315

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

SciTech Connect

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

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

1988-10-01T23:59:59.000Z

316

UNBC SAFETY CHECKLIST SAFETY CHECKLIST  

E-Print Network (OSTI)

1 UNBC SAFETY CHECKLIST SAFETY CHECKLIST INSTRUCTIONS PAGE Please use the following table below needs, contact the Risk & Safety Department at 250-960- (5530) for further instructions. This safety. The safety checklist also helps you to establish due diligence under Federal and Provincial safety laws

Northern British Columbia, University of

317

Toolbox Safety Talk Ladder Safety  

E-Print Network (OSTI)

Toolbox Safety Talk Ladder Safety Environmental Health & Safety Facilities Safety & Health Section Health & Safety for recordkeeping. Slips, trips, and falls constitute the majority of general industry elevated work tasks. Like any tool, ladders must be used properly to ensure employee safety. GENERAL

Pawlowski, Wojtek

318

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

319

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

E-Print Network (OSTI)

of the fixed-bed reactor, among other disadvantages, is that the reaction is very exothermic, which is a concern in terms of safety hazards and also in terms of cost of heat removal. With the slurry reactor, a problem is that in the liquid media... at different lengths.4 After the reaction takes place, the amount of carbon monoxide consumed decreases and carbon dioxide is produced as a side product.9 The FTS reaction is an extremely exothermic process, which represents serious challenges...

Hamad, Natalie

2012-02-14T23:59:59.000Z

320

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

SciTech Connect

At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRCs assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978 and 1988. All of the reports reviewed were the regulators assessment of the PSR findings rather than the original PSR report, and all but one were English translations from the original language. In these reviews, it was found that most of the countries base their regulatory guidance to some extent (and often to a large extent) on U.S. design codes and standards, NRC regulatory guidance, and U.S. industry guidance. In addition, many of the observed operational technical issues and OpE events reported for U.S. reactors are also cited in the PSR reports. The PSR reports also identified a number of potential technical material/component performance issues and OpE events that are not commonly reported for U.S. plants.

Chopra, Omesh K. [Argonne National Lab., IL (United States). Environmental Science Division; Diercks, Dwight R. [Argonne National Lab., IL (United States). Nuclear Engineering Division; Ma, David Chia-Chiun [Argonne National Lab., IL (United States). Environmental Science Division; Garud, Yogendra S. [Argonne National Lab., IL (United States). Environmental Science Division

2013-12-17T23:59:59.000Z

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

Analysis of higher order optical aberrations in the SLC final focus using Lie Algebra techniques  

SciTech Connect

The SLC final focus system is designed to have an overall demagnification of 30:1, with a {beta} at the interaction point ({beta}*) of 5 mm, and an energy band pass of {approximately}0.4%. Strong sextupole pairs are used to cancel the large chromaticity which accrues primarily from the final triplet. Third-order aberrations limit the performance of the system, the dominating terms being U{sub 1266} and U{sub 3466} terms (in the notation of K. Brown). Using Lie Algebra techniques, it is possible to analytically calculate the soave of these terms in addition to understanding their origin. Analytical calculations (using Lie Algebra packages developed in the Mathematica language) are presented of the bandwidth and minimum spot size as a function of divergence at the interaction point (IP). Comparisons of the analytical results from the Lie Algebra maps and results from particle tracking (TURTLE) are also presented.

Walker, N.J.; Irwin, J.; Woodley, M.

1993-04-01T23:59:59.000Z

322

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

323

Combining Formal Methods and Safety Analysis -The ForMoSA Approach  

E-Print Network (OSTI)

techniques [13, 20] like failure modes and effects analysis (FMEA) or fault tree analysis (FTA). The combina in which a component may fail. The leaves of all fault trees are failure modes. The starting column of FMEA

Reif, Wolfgang

324

Formal Support for Quantitative Analysis of Residual Risks in Safety-Critical Systems  

E-Print Network (OSTI)

and Effects Analysis (FMEA) and Fault-Tree Analy- sis (FTA) [16]. However, many of these techniques become

325

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

SciTech Connect

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

326

Safety First Safety Last Safety Always Requirements for employers  

E-Print Network (OSTI)

Safety First Safety Last Safety Always Requirements for employers · Fallprotectionsandproperuseofrelated-safety equipmentsuchaslifelines,harness · Properuseofdangeroustools,thenecessaryprecautionstotake,andtheuseof theprotectiveandemergencyequipmentrequired. Safety Training and Education Safety Tip #18 Get smart. Use safety from the start. All

Minnesota, University of

327

POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHealth and Safety Impacts of Nuclear, Geothermal, and Fossil- FuelHealth and Safety Aspects of Pro- posed Nuclear, Geothermal, and Fossil-Fuel

Nero, A.V.

2010-01-01T23:59:59.000Z

328

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHealth and Safety Impacts of Nuclear, Geothermal, and Fossil- Fuel

Nero, jA.V.

2010-01-01T23:59:59.000Z

329

A REVIEW OF AIR QUALITY MODELING TECHNIQUES. VOLUME 8 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHealth and Safety Impacts of Nuclear, Geothermal, and Fossil- Fuel

Rosen, L.C.

2010-01-01T23:59:59.000Z

330

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUELHealth and Safety Impacts of Nuclear, Geothermal, and Fossil- FuelHealth and Safety Aspects of Pro- posed Nuclear, Geothermal, and Fossil-Fuel

Yen, W.W.S.

2010-01-01T23:59:59.000Z

331

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)  

SciTech Connect

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

332

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)

333

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

334

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

335

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 1, Plenary sessions, reactor licensing topics, NUREG-1150, risk analysis/PRA applications, innovative concepts for increased safety of advanced power reactors, severe accident modeling and analysis  

SciTech Connect

This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 1, discusses the following: plenary sessions; reactor licensing; NUREG-1150; risk analysis; innovative concepts for increased safety of advanced power reactors; and severe accident modeling and analysis. Thirty-two reports have been cataloged separately.

Weiss, A.J. (comp.)

1988-02-01T23:59:59.000Z

336

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

SciTech Connect

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

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

1988-07-01T23:59:59.000Z

337

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

SciTech Connect

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

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

1988-07-01T23:59:59.000Z

338

On the Partial-Wave Analysis of Mesonic Resonances Decaying to Multiparticle Final States Produced by Polarized Photons  

SciTech Connect

Meson spectroscopy is going through a revival with the advent of high statistics experiments and new advances in the theoretical predictions. The Constituent Quark Model (CQM) is finally being expanded considering more basic principles of field theory and using discrete calculations of Quantum Chromodynamics (lattice QCD). These new calculations are approaching predictive power for the spectrum of hadronic resonances and decay modes. It will be the task of the new experiments to extract the meson spectrum from the data and compare with those predictions. The goal of this report is to describe one particular technique for extracting resonance information from multiparticle final states. The technique described here, partial wave analysis based on the helicity formalism, has been used at Brookhaven National Laboratory (BNL) using pion beams, and Jefferson Laboratory (Jlab) using photon beams. In particular this report broaden this technique to include production experiments using linearly polarized real photons or quasi-real photons. This article is of a didactical nature. We describe the process of analysis, detailing assumptions and formalisms, and is directed towards people interested in starting partial wave analysis.

Salgado, Carlos W. [Norfolk State University, Norfolk, VA (United States) and Thomas Jefferson National Accelerator Facility, Newport News, VA (United States); Weygand, Dennis P. [Thomas Jefferson National Accelerator Facility, Newport News, VA (United States)

2014-04-01T23:59:59.000Z

339

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

SciTech Connect

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

340

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;

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

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;

342

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

SciTech Connect

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

343

UNBC SAFETY CHECKLIST SAFETY CHECKLIST  

E-Print Network (OSTI)

1 UNBC SAFETY CHECKLIST SAFETY CHECKLIST INSTRUCTIONS PAGE Please use the following table below needs, contact the Risk & Safety Department at 250-960- (5530) for further instructions. This safety to remain safe here at UNBC. The safety checklist also helps you to establish due diligence under Federal

Northern British Columbia, University of

344

Integrated Safety Management Safety Culture Resources | Department...  

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

Safety Culture Resources Integrated Safety Management Safety Culture Resources A collection of resources available in implementing ISM safety culture activities Safety from the...

345

An interdisciplinary mathematical approach to the analysis and development. Final report  

SciTech Connect

Brief summaries are given for the work done in the following three areas: (1) noninvertible dynamical systems; (2) equation symmetries and boundary layer flows; and (3) intermediate models and data analysis.

Curry, J.H.

1998-08-01T23:59:59.000Z

346

Formal Safety analysis of a radio-based railroad crossing using Deductive Cause-Consequence  

E-Print Network (OSTI)

and effects analysis (FMEA) and fault tree analysis (FTA). We apply the method to a real world case study like FMEA [10], FMECA [4] and FTA [3]. The logical framework of DCCA may be used to rigorously verify of what can by analyzed) than traditional FMEA. We show, that the results of DCCA have the same semantics

Reif, Wolfgang

347

Safety Analysis of an Airbag System Using Probabilistic FMEA and Probabilistic Counterexamples  

Science Journals Connector (OSTI)

Failure mode and effects analysis (FMEA) isa technique to reason about possible system hazards thatresult from system or system component failures. Traditionally, FMEA does not take the probabilities with which these failures may occur into account. ... Keywords: Probabilistic FMEA, Stochastic Modelling, Stochastic Model Checking, Counter Examples in Stochastic Model Checking, Dependability Analysis

Husain Aljazzar; Manuel Fischer; Lars Grunske; Matthias Kuntz; Florian Leitner-Fischer; Stefan Leue

2009-09-01T23:59:59.000Z

348

Reliability analysis of safety-related digitalinstrumentation and controlin a nuclear power plant.  

E-Print Network (OSTI)

?? There is so far no consensus on how to develop a reliability model of safety-related digitalinstrumentation and control (I&C) in a probabilistic safety assessment (more)

Gustafsson, Johan

2012-01-01T23:59:59.000Z

349

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

SciTech Connect

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

350

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

351

Toolbox Safety Talk Safety Data Sheets (SDS)  

E-Print Network (OSTI)

Toolbox Safety Talk Safety Data Sheets (SDS) Environmental Health & Safety Facilities Safety-in sheet to Environmental Health & Safety for recordkeeping. Chemical manufacturers are required to produce Safety Data Sheets (SDS) for all chemicals produced. "Safety Data Sheets", previously referred

Pawlowski, Wojtek

352

Public Safety Public Safety Center  

E-Print Network (OSTI)

and bring someone with you or visit a grocery store or gas station. Personal Safety Precautions Safety the police, or go di- rectly to the police station or Public Safety. Do not label keys with your name or any

353

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

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

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

354

2012 Annual Workforce Analysis and Staffing Plan Report- Office of Health, Safety and Security  

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

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

355

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

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

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

356

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

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

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

357

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

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

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

358

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

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

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

359

Preliminary Study on Reliability Analysis of Safety I&C System in NPP  

Science Journals Connector (OSTI)

Digital instrumentation and control (I&C) systems, such as digital Reactor Protection System (RPS), are being employed ... upgraded Nuclear Power Plants (NPPs). The reliability analysis of digital I&C system turn...

Chao Guo; Duo Li; Huasheng Xiong

2012-01-01T23:59:59.000Z

360

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

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

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

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

2014 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety  

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

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

362

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

SciTech Connect

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

363

TE-2013-000393 Final 1 Abstract--Learning the analysis of electrical circuits  

E-Print Network (OSTI)

related to resistance. The color group achieved significantly higher post-test scores, gave higher ratings.edu. Elementary electrical circuit analysis mainly revolves around the three quantities related by Ohm's Law, namely voltage, current, and resistance. These three quantities typically appear throughout the multiple

Reisslein, Martin

364

Health Safety and Environmental Protection Page 1  

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

Meeting Summary March 10, 2011 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE March 10, 2011 Richland, WA Topics in this...

365

Cryogenic safety  

Science Journals Connector (OSTI)

Cryogenic safety ... Examines the properties of cryogenic fluids and hazards associated with their use and storage. ...

Eric W. Spencer

1964-01-01T23:59:59.000Z

366

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

367

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

368

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

369

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

370

Benefits analysis for the production of fuels and chemicals using solar thermal energy. Final report  

SciTech Connect

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

371

Electromagnetic transients program (EMTP): Volume 4, Workbook IV (TACS) (Transients Analysis of Control Systems): Final report  

SciTech Connect

This workbook represents an introduction to the use of TACS (Transients Analysis of Control Systems) in the EMTP. The material progresses from an overview of basic TACS concepts and components to a detailed HVDC model. The following application of TACS are covered: a variable load problem, static Var systems, thyristor models, TCR, basic HVDC models and a detailed HVDC model. Complete data files are given for most examples.

Lasseter, R.H.

1989-06-01T23:59:59.000Z

372

Recovery of ethylene glycol from used antifreeze, Phase 1. Market analysis. Final report  

SciTech Connect

Objective is to conduct a market analysis on the recovery of ethylene glycol from used antifreeze. The report begins with a theoretical analysis of various recovery processes. An estimate of the annual energy savings from an ethylene glycol recycle industry is made. The basic manufacturing process is discussed and the total production energy requirement is shown to be approximately 15,000 BTU per pound ethylene glycol. The plant used in the baseline case processes 400,000 gallons of used antifreeze and yields 1.82 million pounds of ethylene glycol per year. This amount of used antifreeze would be available for recycle from service stations (at a price of $0.05 per pound) from an area with a population of one million inhabitants. Laboratory analyses of used antifreeze samples were made. Tests were performed to determine if oil and antifreeze could be collected as a mixture and then separated during the recovery process. Antifreeze production levels in the United States are typically around 200 million gallson per year. There is a willingness to recycle antifreeze if the price is high enough. The economic analysis of the recycle process indicates a total plant cost of $448,200 and an initial investment of $633,268. The case of combined oil and antifreeze is considered.

Strand, M.J.; Uvelli, D.A.

1986-09-01T23:59:59.000Z

373

ANALYSIS OF SAFETY RELIEF VALVE PROOF TEST DATA TO OPTIMIZE LIFECYCLE MAINTENANCE COSTS  

SciTech Connect

Proof test results were analyzed and compared with a proposed life cycle curve or hazard function and the limit of useful life. Relief valve proof testing procedures, statistical modeling, data collection processes, and time-in-service trends are presented. The resulting analysis of test data allows for the estimation of the PFD. Extended maintenance intervals to the limit of useful life as well as methodologies and practices for improving relief valve performance and reliability are discussed. A generic cost-benefit analysis and an expected life cycle cost reduction concludes that $90 million maintenance dollars might be avoided for a population of 3000 valves over 20 years.

Gross, Robert; Harris, Stephen

2007-08-01T23:59:59.000Z

374

Software system reliability and safety assessment: an extended FMEA approach  

Science Journals Connector (OSTI)

This paper presents a methodology for assessing the reliability and safety of a software based on an extended Failure Modes and Effects Analysis (FMEA) approach. The methodology is described in steps with illustrative examples. The analysis starts from initial phase of the software development and evolves during the subsequent phases of software development providing valuable information to each phases. Finally, the analysis yields a quantitative assessment of reliability and safety of the software system. The paper's main objective is to support Probabilistic Safety Assessment (PSA) in assessing risk. Risk is a function of severity and failure frequency/probability. The severity is characteristic of failure effects. Failures may be analysed as functional or component failure. In this paper, it is proposed to consider severity levels at functional failure level as it is easier to understand failure effects at functional level. Moreover, various logical combinations of different functional failures can also be formed and analysed using the proposed approach.

Sinda Rebello; Neeraj Kumar Goyal

2010-01-01T23:59:59.000Z

375

Facility Safety  

Directives, Delegations, and Requirements

The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.

2000-11-20T23:59:59.000Z

376

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

377

On the Safety of Nocker's Strictness Analysis Manfred Schmidt-Schau1  

E-Print Network (OSTI)

Frankfurt, Germany, schauss@ki.informatik.uni-frankfurt.de 2 Dept. of Mathematics and Computing Science Intelligence and Software Technology, Institut f¨ur Informatik, J.W.Goethe-Universit¨at Frankfurt, 30.10.2004 Abstract. This paper proves correctness of N¨ocker's method of strict- ness analysis, implemented for Clean

Schmidt-Schauss, Manfred

378

Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety  

SciTech Connect

This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

1993-03-01T23:59:59.000Z

379

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

SciTech Connect

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

380

METHODOLOGIES FOR REVIEW OF THE HEALTH AND SAFETY ASPECTS OF PROPOSED NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL SITES AND FACILITIES. VOLUME 9 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

of the health and safety impact of fossil fuel emissions.to public health and safety, of any fossil fuel plant areHEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL

Nero, A.V.

2010-01-01T23:59:59.000Z

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381

Proceedings of the US Nuclear Regulatory Commission fourteenth water reactor safety information meeting: Volume 1, Plenary session, Severe accident sequence analysis, Risk analysis/PRA applications, Reference plant risk analysis - NUREG-1150, Innovative concepts for increased safety of advanced power reactors  

SciTech Connect

This six-volume report contains 156 papers out of the 175 that were presented at the Fourteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 27-31, 1986. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-four different papers presented by researchers from Canada, Czechoslovakia, Finland, Germany, Italy, Japan, Mexico, Spain, Sweden, Switzerland and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

382

Analysis of Chinook Salmon in the Columbia River from an Ecosystem Perspective. Final Report.  

SciTech Connect

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

383

Tools for Accurate and Efficient Analysis of Complex Evolutionary Mechanisms in Microbial Genomes. Final Report  

SciTech Connect

I proposed to develop computationally efficient tools for accurate detection and reconstruction of microbes' complex evolutionary mechanisms, thus enabling rapid and accurate annotation, analysis and understanding of their genomes. To achieve this goal, I proposed to address three aspects. (1) Mathematical modeling. A major challenge facing the accurate detection of HGT is that of distinguishing between these two events on the one hand and other events that have similar "effects." I proposed to develop a novel mathematical approach for distinguishing among these events. Further, I proposed to develop a set of novel optimization criteria for the evolutionary analysis of microbial genomes in the presence of these complex evolutionary events. (2) Algorithm design. In this aspect of the project, I proposed to develop an array of e#14;cient and accurate algorithms for analyzing microbial genomes based on the formulated optimization criteria. Further, I proposed to test the viability of the criteria and the accuracy of the algorithms in an experimental setting using both synthetic as well as biological data. (3) Software development. I proposed the #12;nal outcome to be a suite of software tools which implements the mathematical models as well as the algorithms developed.

Nakhleh, Luay

2014-03-12T23:59:59.000Z

384

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

385

Post-test analysis of dryout test 7B' of the W-1 Sodium Loop Safety Facility Experiment with the SABRE-2P code. [LMFBR  

SciTech Connect

An understanding of conditions that may cause sodium boiling and boiling propagation that may lead to dryout and fuel failure is crucial in liquid-metal fast-breeder reactor safety. In this study, the SABRE-2P subchannel analysis code has been used to analyze the ultimate transient of the in-core W-1 Sodium Loop Safety Facility experiment. This code has a 3-D simple nondynamic boiling model which is able to predict the flow instability which caused dryout. In other analyses dryout has been predicted for out-of-core test bundles and so this study provides additional confirmation of the model.

Rose, S.D.; Dearing, J.F.

1981-01-01T23:59:59.000Z

386

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

387

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

388

Technical and economic analysis: Gas cofiring in industrial boilers. Final report, November 1995-September 1996  

SciTech Connect

This report presents an analysis of the technical and marketing issues associated with the deployment of natural gas cofiring technology in stoker boilers. As part of the work effort, a composite database of stoker boilers was developed using state and federal emission inventories over the years 1985 - 1995. Information sources included the most recent AIRS Facility Subsystem database, the Ozone Transport Region 1990 database, the 1990 Ohio Permit database and the 1985 NAPAP database--all are electronic databases of facilities with air emission permits. The initial data set included almost 3,000 stokers at about 1,500 locations. Stoker facilities were contacted to verify the operating status, capacity, fuel capability, efficiency and other stoker-specific data. The report presents the current stoker boiler distribution by SIC, industrial groups, primary solid fuel (coal, wood, waste, refuse), operating status, and state. Maps are included.

Potter, F.J.

1996-09-01T23:59:59.000Z

389

Analysis of Devonian shale multiwell interference tests in Meigs County, Ohio. Final report  

SciTech Connect

The Offset Well Test Program completed in 1981 was undertaken in order to investigate the production characteristics of Devonian shale reservoirs. The investigation involved a study of gas flow through natural fractures, the orientation and distribution of these fractures, and the gas storage/release mechanism and its effect on production. An experiment was designed to test the reservoir under strictly controlled conditions. Interference tests were conducted in Meigs County, Ohio on two wells drilled in the expected maximum and minimum permeability directions from a producing well with known completion and production history. Analysis of the test results indicate that the Devonian shale formation in the Meigs County, Ohio, area is an anisotropic, layered reservoir system. Flow characteristics indicate that the Meigs County reservoir is naturally fractured, may be represented as a dual porosity system, and may be modeled using pseudo-steady-state gas transfer from the matrix to the fracture system. The orientation of the natural fracture system was established through core observation and well test analysis as S65/sup 0/W. The maximum to minimum permeability ratio in the direction of the natural fracture system was calculated to be 8.3. Three distinct zones with independent flow characteristics were identified. The bottom zone, with permeability values significantly higher than the upper two zones, is highly fractured and is a major contributor to the gas production of Well 10056. The pressure profiles of the bottom zones relative to the upper zones were significantly different, indicating minimal communication between the layers. The knowledge of these parameters should have a significant impact on future development of shale reservoirs, through optimization of well spacing and choice of stimulation treatment to enhance gas production.

Alam, J.; Horan, K.; Lee, B.; Sawyer, W.

1982-02-23T23:59:59.000Z

390

SOLERAS - Solar Energy Water Desalination Project: Catalytic. System design final report. Volume 1. System requirement definition and system analysis  

SciTech Connect

Catalytic Inc. was awarded a contract to conduct a preliminary system design and cost analysis for a brackish water desalination project to be located in Brownsville, Texas. System analyses and economic analyses were performed to define the baseline solar energy desalination system. The baseline system provides an average daily product water capacity of 6000 mT. The baseline system is optimal relative to technological risk, performance, and product water cost. Subsystems and their interfaces have been defined and product water cost projections made for commercial plants in a range of capacities. Science Applications, Inc. (SAI), subcontractor to Catalytic, had responsibility for the solar power system. This, the final report, summarizes the work performed under the Phase I effort.

Not Available

1986-01-01T23:59:59.000Z

391

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

SciTech Connect

The RiskMod Model is comprised of a set of risk simulation models, collectively referred to as RiskSim; a set of computer programs that manages data referred to as Data Management Procedures; and RevSim, a model that calculates net revenues. RiskMod interacts with the AURORA Model, the RAM2007, and the ToolKit Model during the process of performing the Risk Analysis Study. AURORA is the computer model being used to perform the Market Price Forecast Study (see Market Price Forecast Study, WP-07-FS-BPA-03); the RAM2007 is the computer model being used to calculate rates (see Wholesale Power Rate Development Study, WP-07-FS-BPA-05); and the ToolKit is the computer model being used to develop the risk mitigation package that achieves BPA's 92.6 percent TPP standard (see Section 3 in the Risk Analysis Study, WP-07-FS-BPA-04). Variations in monthly loads, resources, natural gas prices, forward market electricity prices, transmission expenses, and aluminum smelter benefit payments are simulated in RiskSim. Monthly spot market electricity prices for the simulated loads, resources, and natural gas prices are estimated by the AURORA Model. Data Management Procedures facilitate the format and movement of data that flow to and/or from RiskSim, AURORA, and RevSim. RevSim estimates net revenues using risk data from RiskSim, spot market electricity prices from AURORA, loads and resources data from the Load Resource Study, WP-07-FS-BPA-01, various revenues from the Revenue Forecast component of the Wholesale Power Rate Development Study, WP-07-FSBPA-05, and rates and expenses from the RAM2007. Annual average surplus energy revenues, purchased power expenses, and section 4(h)(10)(C) credits calculated by RevSim are used in the Revenue Forecast and the RAM2007. Heavy Load Hour (HLH) and Light Load Hour (LLH) surplus and deficit energy values from RevSim are used in the Transmission Expense Risk Model. Net revenues estimated for each simulation by RevSim are input into the ToolKit Model to develop the risk mitigation package that achieves BPA's 92.6 percent TPP standard. The processes and interaction between each of the models and studies are depicted in Graph 1.

United States. Bonneville Power Administration.

2006-07-01T23:59:59.000Z

392

Water-lithium bromide double-effect absorption cooling analysis. Final report  

SciTech Connect

This investigation involved the development of a numerical model for the transient simulation of the double-effect, water-lithium bromide absorption cooling machine, and the use of the model to determine the effect of the various design and input variables on the absorption unit performance. The performance parameters considered were coefficient of performance and cooling capacity. The sensitivity analysis was performed by selecting a nominal condition and determining performance sensitivity for each variable with others held constant. The variables considered in the study include source hot water, cooling water, and chilled water temperatures; source hot water, cooling water, and chilled water flow rates; solution circulation rate; heat exchanger areas; pressure drop between evaporator and absorber; solution pump characteristics; and refrigerant flow control methods. The performance sensitivity study indicated in particular that the distribution of heat exchanger area among the various (seven) heat exchange components is a very important design consideration. Moreover, it indicated that the method of flow control of the first effect refrigerant vapor through the second effect is a critical design feature when absorption units operate over a significant range of cooling capacity. The model was used to predict the performance of the Trane absorption unit with fairly good accuracy. The dynamic model should be valuable as a design tool for developing new absorption machines or modifying current machines to make them optimal based on current and future energy costs.

Vliet, G.C.; Lawson, M.B.; Lithgow, R.A.

1980-12-01T23:59:59.000Z

393

Major models and data sources for residential and commercial sector energy conservation analysis. Final report  

SciTech Connect

Major models and data sources are reviewed that can be used for energy-conservation analysis in the residential and commercial sectors to provide an introduction to the information that can or is available to DOE in order to further its efforts in analyzing and quantifying their policy and program requirements. Models and data sources examined in the residential sector are: ORNL Residential Energy Model; BECOM; NEPOOL; MATH/CHRDS; NIECS; Energy Consumption Data Base: Household Sector; Patterns of Energy Use by Electrical Appliances Data Base; Annual Housing Survey; 1970 Census of Housing; AIA Research Corporation Data Base; RECS; Solar Market Development Model; and ORNL Buildings Energy Use Data Book. Models and data sources examined in the commercial sector are: ORNL Commercial Sector Model of Energy Demand; BECOM; NEPOOL; Energy Consumption Data Base: Commercial Sector; F.W. Dodge Data Base; NFIB Energy Report for Small Businesses; ADL Commercial Sector Energy Use Data Base; AIA Research Corporation Data Base; Nonresidential Buildings Surveys of Energy Consumption; General Electric Co: Commercial Sector Data Base; The BOMA Commercial Sector Data Base; The Tishman-Syska and Hennessy Data Base; The NEMA Commercial Sector Data Base; ORNL Buildings Energy Use Data Book; and Solar Market Development Model. Purpose; basis for model structure; policy variables and parameters; level of regional, sectoral, and fuels detail; outputs; input requirements; sources of data; computer accessibility and requirements; and a bibliography are provided for each model and data source.

Not Available

1980-09-01T23:59:59.000Z

394

Lower Columbia River and Estuary Ecosystem Restoration Program Reference Site Study: 2011 Restoration Analysis - FINAL REPORT  

SciTech Connect

The Reference Site (RS) study is part of the research, monitoring, and evaluation (RME) effort developed by the Action Agencies (Bonneville Power Administration [BPA], U.S. Army Corps of Engineers, Portland District [USACE], and U.S. Bureau of Reclamation) in response to Federal Columbia River Power System (FCRPS) Biological Opinions (BiOp). While the RS study was initiated in 2007, data have been collected at relatively undisturbed reference wetland sites in the LCRE by PNNL and collaborators since 2005. These data on habitat structural metrics were previously summarized to provide baseline characterization of 51 wetlands throughout the estuarine and tidal freshwater portions of the 235-km LCRE; however, further analysis of these data has been limited. Therefore, in 2011, we conducted additional analyses of existing field data previously collected for the Columbia Estuary Ecosystem Restoration Program (CEERP) - including data collected by PNNL and others - to help inform the multi-agency restoration planning and ecosystem management work underway in the LCRE.

Borde, Amy B.; Cullinan, Valerie I.; Diefenderfer, Heida L.; Thom, Ronald M.; Kaufmann, Ronald M.; Zimmerman, Shon A.; Sagar, Jina; Buenau, Kate E.; Corbett, C.

2012-05-31T23:59:59.000Z

395

Model-based safety assessments  

SciTech Connect

Sandia National Laboratories performs systems analysis of high risk, high consequence systems. In particular, Sandia is responsible for the engineering of nuclear weapons, exclusive of the explosive physics package. In meeting this responsibility, Sandia has developed fundamental approaches to safety and a process for evaluating safety based on modeling and simulation. These approaches provide confidence in the safety of our nuclear weapons. Similar concepts may be applied to improve the safety of other high consequence systems.

Carlson, D.D.; Jones, T.R.

1998-04-01T23:59:59.000Z

396

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

397

Facility Safety  

Directives, Delegations, and Requirements

The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.

2005-12-22T23:59:59.000Z

398

Biological Safety  

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

The DOE's Biological Safety Program provides a forum for the exchange of best practices, lessons learned, and guidance in the area of biological safety. This content is supported by the Biosurety Executive Team. The Biosurety Executive Team is a DOE-chartered group. The DOE Office of Worker Safety and Health Policy provides administrative support for this group. The group identifies biological safety-related issues of concern to the DOE and pursues solutions to issues identified.

399

Safety analysis of exothermic reaction hazards associated with the organic liquid layer in tank 241-C-103  

SciTech Connect

Safety hazards associated with the interim storage of a potentially flammable organic liquid in waste Tank C-103 are identified and evaluated. The technical basis for closing the unreviewed safety question (USQ) associated with the floating liquid organic layer in this tank is presented.

Postma, A.K.; Bechtold, D.B.; Borsheim, G.L.; Grisby, J.M.; Guthrie, R.L.; Kummerer, M.; Turner, D.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

1994-03-01T23:59:59.000Z

400

Safety Information for Families  

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

Safety Information for Families Checking your home for hazards 22 safety items no home should be without Home Safety Checklists Helpful links Home Safety Council Hunter Safety:...

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

Facility Safety  

Directives, Delegations, and Requirements

Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.

1995-10-13T23:59:59.000Z

402

Facility Safety  

Directives, Delegations, and Requirements

The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.

2012-12-04T23:59:59.000Z

403

Facility Safety  

Directives, Delegations, and Requirements

This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.

2005-12-22T23:59:59.000Z

404

SU?C?BRCD?06: A Method of Streamlined Failure Mode and Effect Analysis to Improve Patient Safety  

Science Journals Connector (OSTI)

Purpose: Failure Mode and Effects Analysis (FMEA) provides a proactive method of improving the quality and safety of treatments by identifying and correcting hazards points in the process of clinical care. FMEA however is often thought to be prohibitively labor intensive. This study outlines a method of streamlined FMEA conducted with limited resources and assesses its feasibility and effectiveness. Methods: FMEA was performed on the external beam service of a clinic treating approximately 650 patients per year on three linear accelerators. A facilitator and local team leader were identified and a plan was developed to complete the exercise in four one? hour meetings as follows: 1 (core group) introduction and process mapping 2 (all staff) identification of failure modes from expert user input 3 (core group) scoring of failure modes according to an FMEA risk priority number RPN i.e. the product of severity occurrence and detectability scores and 4 (all staff) error proofing of the top?five ranked failure modes. Results: Fifty? two failure modes were identified 43 of which were scored and ranked. Specific interventions were developed for the five highest ranked failure modes. FMEA scoring after intervention indicated that the average RPN score for the top five modes decreased from 273 to 161 (p=0.03) while FMEA scoring of a control group of failure modes with no intervention did not show a significant change in RPN (p=0.07). The exercise was accomplished within the expected timeline and required 55 total hours of staff time and 20 hours of facilitator effort. Conclusion: Streamlined FMEA analysis is feasible with a relatively modest effort and can reduce the risk profile of a facility. This may open the possibility of performing repeat FMEA on a regular basis.

E Ford; K Smith; J Keck; K Harris; S Terezakis; G Sibley

2012-01-01T23:59:59.000Z

405

Tank 241-BY-109, cores 201 and 203, analytical results for the final report  

SciTech Connect

This document is the final laboratory report for tank 241-BY-109 push mode core segments collected between June 6, 1997 and June 17, 1997. The segments were subsampled and analyzed in accordance with the Tank Push Mode Core Sampling and Analysis Plan (Bell, 1997), the Tank Safety Screening Data Quality Objective (Dukelow, et al, 1995). The analytical results are included.

Esch, R.A.

1997-11-20T23:59:59.000Z

406

The 2dF Galaxy Redshift Survey: Power-spectrum analysis of the final dataset and cosmological implications  

E-Print Network (OSTI)

We present a power spectrum analysis of the final 2dF Galaxy Redshift Survey, employing a direct Fourier method. The sample used comprises 221,414 galaxies with measured redshifts. We investigate in detail the modelling of the sample selection. A new angular mask is derived, based on revisions to the photometric calibration. The redshift selection function is determined by dividing the survey according to rest-frame colour, and deducing a self-consistent treatment of k-corrections and evolution for each population. The covariance matrix for the power-spectrum estimates is determined using two different approaches to the construction of mock surveys which are used to demonstrate that the input cosmological model can be correctly recovered. We are confident that the 2dFGRS power spectrum can be used to infer the matter content of the universe. On large scales, our estimated power spectrum shows evidence for the `baryon oscillations' that are predicted in CDM models. Fitting to a CDM model, assuming a primordial $n_{s}=1$ spectrum, $h=0.72$ and negligible neutrino mass, the preferred parameters are $\\Omega_{M} h = 0.168 \\pm 0.016$ and a baryon fraction $\\Omega_{b} /\\Omega_{M} = 0.185\\pm0.046$ (1$\\sigma$ errors). The value of $\\Omega_{M} h$ is $1\\sigma$ lower than the $0.20 \\pm 0.03$ in our 2001 analysis of the partially complete 2dFGRS. This shift is largely due to the signal from the newly-sampled regions of space, rather than the refinements in the treatment of observational selection. This analysis therefore implies a density significantly below the standard $\\Omega_{M} =0.3$: in combination with CMB data from WMAP, we infer $\\Omega_{M} =0.231\\pm 0.021$. (Abridged.)

S. Cole; W. J. Percival; J. A. Peacock; P. Norberg; C. M. Baugh; C. S. Frenk; I. Baldry; J. Bland-Hawthorn; T. Bridges; R. Cannon; M. Colless; C. Collins; W. Couch; N. J. G. Cross; G. Dalton; V. R. Eke; R. De Propris; S. P. Driver; G. Efstathiou; R. S. Ellis; K. Glazebrook; C. Jackson; A. Jenkins; O. Lahav; I. Lewis; S. Lumsden; S. Maddox; D. Madgwick; B. A. Peterson; W. Sutherland; K. Taylor

2005-08-05T23:59:59.000Z

407

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

408

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

409

Safety Cinema: Safety Videos: Los Alamos National Laboratory  

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

Safety CinemaTM VideosINDUSTRIAL HYGIENE AND SAFETY Safety Videos Safety Cinema Safety Videos Home Safety Cinema Human Beings Beryllium Integrated Safety CONTACTS Occupational...

410

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

411

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11  

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

During the 580th meeting of the Advisory Committee on Reactor Safeguards (ACRS), February10-12, 2011, we reviewed the staffs white paper, A Comparison of Integrated Safety Analysisand...

412

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

E-Print Network (OSTI)

involved. In 2009, President Barack Obama (2009) expressed the feelings of many American parents, ?In the end, food safety is something I take seriously, not just as your President, but as a parent. When I heard peanut products were being contaminated... involved. In 2009, President Barack Obama (2009) expressed the feelings of many American parents, ?In the end, food safety is something I take seriously, not just as your President, but as a parent. When I heard peanut products were being contaminated...

Joy, Nathaniel Allen

2012-02-14T23:59:59.000Z

413

Energy Engineering Analysis Program for the 193rd Infantry Brigade (Panama). Executive summary, phase III. Final report  

SciTech Connect

A 14-month long Energy Engineering Analysis Program (EEAP) has been conducted for the 193d Infantry Brigade. The EEAP identified nine ECIP Projects (Increment A and Increment B), four non-ECIP Projects (Increment G) and fourteen Facility Engineer Energy Conservation Measures (Increment F). Five of the nine ECIP Projects were expedited for submission in June 1983 for funding in FY 86. One of the ECIP Projects is for DODDS. Twenty other Energy Conservation Measures were investigated, but not recommended. A two-volume Final Report was prepared in October 1983 and six copies delivered to the Installation. Two copies went to FORSCOM and other copies to the Corps Panama Area Engineer, District Engineer, Huntsville Division Engineer, South Atlantic Division Engineer and others. Volume I is Narrative and Supporting Data and Volume II is Project Documentation. For the 15 recommended Projects, complete programming documentation has been prepared. If all 13 were implemented, the Brigade`s total energy consumption (electricity plus DFM) would be reduced by 14 percent and the Brigade would save about $3,200,000 a year. Implementation of the Increment F items would save an additional $164,000 yearly. A prioritized list of energy conservation measures appears later in this Executive Summary. It is important to note that many of the recommended energy conservation measures also qualify for PCIP and QRIP funding and should be simultaneously submitted for funding under these programs.

NONE

1983-09-01T23:59:59.000Z

414

Nuclear Safety Analysis Reports  

Directives, Delegations, and Requirements

Cancels DOE O 5481.1B; paragraphs 7b(3), 7e(3) & 8c of DOE O 5480.6; and 51, 7b(3), 7b(4), 7e(3), 8a & 8h of DOE O 5480.5.

1992-04-30T23:59:59.000Z

415

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

416

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

417

Safety Values  

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

* Work-related injuries, illnesses and environmental incidents are preventable. * A just culture exists where safety and environmental concerns are brought forward without fear of...

418

Safety Engineer  

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

This position is located within the Savannah River Operations Office, Office of Safety and Quality Assurance, Technical Support Division. Department of Energy (DOE) Savannah River (SR) Operations...

419

Code of Federal Regulations Occupational Radiation Protection; Final Rule |  

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

Code of Federal Regulations Occupational Radiation Protection; Code of Federal Regulations Occupational Radiation Protection; Final Rule Code of Federal Regulations Occupational Radiation Protection; Final Rule The Department of Energy (DOE) is amending its primary standards for occupational radiation protection. This final rule is the culmination of a systematic analysis to identify the elements of a comprehensive radiation protection program and determine those elements of such a program that should be codified as DOE continues its transition from a system of contractually-based nuclear safety standards to regulatory based requirements. Code of Federal Regulations Occupational Radiation Protection; Final Rule More Documents & Publications Code of Federal Regulations PART 835-OCCUPATIONAL RADIATION PROTECTION Order Module--NNSA OCCUPATIONAL RADIATION PROTECTION

420

EPA Final Ground Water Rule  

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

Office of Nuclear Safety and Environment Office of Nuclear Safety and Environment Nuclear Safety and Environment Information Brief HS-20-IB-2007-02 (March 2007) EPA Final Ground Water Rule Safe Drinking Water Act: National Primary Drinking Water Regulations Ground Water Rule - 40 CFR Parts 9, 141 and 142 Final Rule: 71 FR 65574 Effective Date: January 8, 2007 1 RULE SYNOPSIS On November 8, 2006, the U.S. Environmental Protection Agency (EPA) published a final Ground Water Rule (GWR) to promote increased protection against microbial pathogens that may be present in public water systems (PWSs) that use ground water sources for their supply (these systems are known as ground water systems). This Rule establishes a risk-targeted approach

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.


421

Technical Review Report for the Model 9977 Safety Analysis Report for Packaging Addendum 1 Justification for DNDO Contents  

SciTech Connect

The Model 9977 Package is currently certified for Content Envelope C.1, {sup 238}Pu Heat Sources, either in Radioisotope Thermoelectric Generator (RTG), or in Food-Pack Can configurations, under Certificate of Compliance (CoC) Certificate Number 9977 and Package Identification Number USA/9977/B(M)F-96 (DOE). Addendum 1, Justification for DNDO Contents,--the Submittal--supplements Revision 2 of the Safety Analysis Report for Packaging for the Model 9977 Package. The Submittal adds five new contents to the Model 9977 Package, Content Envelopes, AC.1 through AC.5. The Content Envelopes are neptunium metal, the beryllium-reflected plutonium ball (BeRP Ball), plutonium/uranium metal, plutonium/uranium metal with enhanced wt% {sup 240}Pu (to 50 wt%), and uranium metal. The last three Content Envelopes are stabilized to DOE-STD-3013. These Content Envelopes will be shipped to the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), where they will reside, and, hence, to off-site locations in support of the Department of Homeland Security (DHS) Domestic Nuclear Detection Office (DNDO). The new certificate will apply to a limited number of Model 9977 Packages. At the same time, the Submittal requests an extension of the periodic maintenance requirements from one (1) year to up to five (5) years using Radio-Frequency Identification (RFID) temperature-monitoring systems to measure the ambient storage temperature in order to ensure that the temperature of the Viton{reg_sign} O-rings for the 6-inch Containment Vessel (6CV) remain less than 200 F. The RFIDs have been developed by Argonne National Laboratory. An on-going surveillance program at the K-Area Materials Storage (KAMS) facility at the Savannah River Site, and an on-going examination of Viton{reg_sign} O-rings from mock Primary Containment Vessels (PCVs) at Savannah River National Laboratory (SRNL) provide the technical justification for the extension of the periodic maintenance interval. Where extended periodic maintenance is desired, the decay heat rate for the Model 9977 Package is limited to 15 watts.

West, M H

2008-12-17T23:59:59.000Z

422

Final Regulatory Impact Review/ Final Environmental Assessment/Initial Regulatory  

E-Print Network (OSTI)

Final Regulatory Impact Review/ Final Environmental Assessment/Initial Regulatory Flexibility................................................................................................. 1 2 REGULATORY IMPACT REVIEW................................................................. 2 2 Analysis Amendment 97 to the Fishery Management Plan for Groundfish of the Bering Sea and Aleutian Islands

423

10 CFR part 851, Workers Safety and Health Program | Department...  

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

CFR part 851, Workers Safety and Health Program 10 CFR part 851, Workers Safety and Health Program February 9, 2006 The Department published a final rule to implement the statutory...

424

Fuzzy Synthetic Evaluation of Gas Station Safety  

Science Journals Connector (OSTI)

Based on the comprehensive analysis of hazard factors and evaluation indexes in gas stations, gas station safety is assessed in a fuzzy synthetic ... comprehensive evaluation, the specific safety level of gas stations

Xiaohua Hao; Xiao Feng

2010-01-01T23:59:59.000Z

425

Future U.S. ITER Safety Studies  

SciTech Connect

With the US re-entering the ITER project, the US safety program has been tasked to address safety issues left unresolved during the US absence over the past five years. As a consequence our current and future US ITER safety studies will focus on validating US safety analysis tools that underpin the ITER safety analysis, refining in-vessel dust and tritium inventory safety limits and developing corresponding dust and tritium removal strategies that will demonstrate compliance with ITER limits without hampering operational flexibility of the machine00.

Petti, D.A.; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory (United States)

2005-05-15T23:59:59.000Z

426

HESSD '98 17 Safety concerns at Ontario Hydro: The need for safety  

E-Print Network (OSTI)

HESSD '98 17 Safety concerns at Ontario Hydro: The need for safety management through incident analysis and safety assessment John D. Lee Battelle Seattle Research Center 4000 NE 41st Street Seattle, WA Engineering University of Toronto benfica@mie.utoronto.ca Safety management and the long-term operation

Lee, John D.

427

DOE/EIS-0236/SA-6 Final Supplement Analysis for Pit Manufacturing Facilities at Los Alamos National Laboratory, Stockpile Stewardship and Management Programmatic Environmental Impact Statement  

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

DATE: REPLY TO ATTN OF: DP-45 (G. Palmer, 6-1785) SUBJECT: DETERMINATION OF THE NEED FOR ADDITIONAL NATIONAL ENVIRONMENTAL POLICY ACT (NEPA) REVIEW TO: Dave Beck, DP-20 As requested in your action memorandum, same subject, I have reviewed the attached Final Supplement Analysis for Pit Manufacturing Facilities at Los Alamos National Laboratory, Stockpile Stewardship and Management Programmatic Environmental Impact Statement, dated August 1999. This analysis was prepared in accordance with 10 CFR 1021.314, contains the comments on the draft Supplement Analysis, dated June 1999, and responds to the comments in Appendix D. Based on my review of the six issues analyzed in the Supplement Analysis, I have determined that none of the information and analysis represent substantial changes to the actions

428

Chapter Eleven - Safety Systems  

Science Journals Connector (OSTI)

Abstract This chapter begins by discussing basic protection concepts related to design, and it considers the development of a hazard tree for an upstream oil and gas facility. The chapter then builds upon the hazard tree to develop a safe process by incorporating findings from a hazards analysis such as a failure mode effect analysis (FMEA) or \\{HAZards\\} \\{OPerability\\} study (HAZOPs). The chapter describes the effects of hydrocarbon releases and how safety devices prevent major accidents from occurring. It further discusses the requirements of API RP 14C, which is a modified FEMA. API RP 14C requires a minimum of two independent layers of protection. This is accomplished through the use of a surface safety system and an emergency support system. The elements of a process safety management system are discussed in detail. The chapter ends by covering all aspects of relief devices and then incorporating them into a relief, vent, or flare system.

Maurice I. Stewart Jr.

2014-01-01T23:59:59.000Z

429

SRTC criticality safety technical review: Phase 1 criticality analysis for the 9972-9975 family of shipping casks: (SRT-CMA-940003)  

SciTech Connect

Review of SRT-CMA-940003, ``Phase I Criticality Analysis For The 9972-9975 Family Of Shipping Casks (U). (SRT-CMA-940003).`` January 22, 1994, has been performed by the SRTC Applied Physics Group. The NCSE is a criticality assessment of the 9972-9975 family of shipping casks. This work is a follow-on of a previous criticality safety evaluation, with the differences between this and the previous evaluation are that now wall tolerances are modeled and more sophisticated analytical methods are applied. The NCSE under review concludes that, with one exception, the previously specified plutonium and uranium mass limits for 9972-9975 family of shipping casks do ensure that WSRC Nuclear Criticality Safety Manual requirements (ref. 1) are satisfied. The one exception is that the plutonium mass limit for the 9974 cask had to be reduced from 4.4 to 4.3 kg. In contrast, the 7.5 kg uranium mass limit for the 9974 cask was raised to 14.5 kg, making the uranium mass identical for all casks in this family. This technical review consisted of an independent check of the methods and models employed, application of ANSI/ANS 8.1 and 8.15, and verification of WSRC Nuclear Criticality Safety Manual procedures.

Rathbun, R.

1994-03-02T23:59:59.000Z

430

Department Safety Representatives Department Safety Representative  

E-Print Network (OSTI)

Department Safety Representatives Overview Department Safety Representative Program/Operations Guidance Document The Department Safety Representative (DSR) serves a very important role with implementation of safety, health, and environmental programs on campus. The role of the DSR is to assist

Pawlowski, Wojtek

431

A fuzzy-based reliability approach to evaluate basic events of fault tree analysis for nuclear power plant probabilistic safety assessment  

Science Journals Connector (OSTI)

Abstract Fault tree analysis has been widely utilized as a tool for nuclear power plant probabilistic safety assessment. This analysis can be completed only if all basic events of the system fault tree have their quantitative failure rates or failure probabilities. However, it is difficult to obtain those failure data due to insufficient data, environment changing or new components. This study proposes a fuzzy-based reliability approach to evaluate basic events of system fault trees whose failure precise probability distributions of their lifetime to failures are not available. It applies the concept of failure possibilities to qualitatively evaluate basic events and the concept of fuzzy sets to quantitatively represent the corresponding failure possibilities. To demonstrate the feasibility and the effectiveness of the proposed approach, the actual basic event failure probabilities collected from the operational experiences of the DavidBesse design of the Babcock and Wilcox reactor protection system fault tree are used to benchmark the failure probabilities generated by the proposed approach. The results confirm that the proposed fuzzy-based reliability approach arises as a suitable alternative for the conventional probabilistic reliability approach when basic events do not have the corresponding quantitative historical failure data for determining their reliability characteristics. Hence, it overcomes the limitation of the conventional fault tree analysis for nuclear power plant probabilistic safety assessment.

Julwan Hendry Purba

2014-01-01T23:59:59.000Z

432

HSS Safety Shares  

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

Safety Shares Safety Shares HSS Safety Shares Home Health, Safety and Security Home HSS Safety Shares 2013 Safety Shares National Weather Service - Lightning Safety General Lightning Safety 7 Important Parts of a Cleaning Label Kitchen Knife Safety Lawn and Garden Tool Hazards Rabies Hearing Loss Winter Driving Tips 2012 Safety Shares Holiday Decoration Safety Tips Countdown to Thanksgiving Holiday Fall Season Safety Tips Slips, Trips and Fall Safety Back To School Safety Tips for Motorists Grills Safety and Cleaning Tips Glass Cookware Safety Water Heater Safety FAQs Root Out Lawn and Garden Tool Hazards First Aid for the Workplace Preventing Colon Cancer Yard Work Safety Yard Work Safety - Part 1 Yard Work Safety - Part 2 High Sodium Risks Heart Risk Stair Safety New Ways To Spot Dangerous Tires

433

Facility Safety  

Directives, Delegations, and Requirements

To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.

2002-05-20T23:59:59.000Z

434

LASER SAFETY POLICY MANUAL ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network (OSTI)

LASER SAFETY POLICY MANUAL ISSUED BY ENVIRONMENTAL HEALTH & SAFETY OFFICE OF RADIOLOGICAL SAFETY and GEORGIA TECH LASER SAFETY COMMITTEE July 1, 2010 Revised July 31, 2012 #12;Laser Safety Program 1-1 #12;Laser Safety Policy Manual TABLE OF CONTENTS 1. POLICY AND SCOPE

Houston, Paul L.

435

Toolbox Safety Talk Machine Shop Safety  

E-Print Network (OSTI)

Toolbox Safety Talk Machine Shop Safety Environmental Health & Safety Facilities Safety & Health to Environmental Health & Safety for recordkeeping. Machine shops are an integral part of the Cornell University be taken seriously. Many of the most frequently cited OSHA safety standards pertain to machine safeguarding

Pawlowski, Wojtek

436

Health, Safety, and Environmental Protection Committee Page 1  

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

May 9, 2013 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE May 9, 2013 Richland, WA Topics in this Meeting Summary Opening...

437

Health, Safety, and Environmental Protection Committee Page 1  

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

August 8, 2013 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY, AND ENVIRONMENTAL PROTECTION COMMITTEE August 8, 2013 Richland, WA Topics in this Meeting Summary...

438

Health Safety and Environmental Protection Committee Page 1  

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

Meeting Summary May 10, 2011 FINAL MEETING SUMMARY HANFORD ADVISORY BOARD HEALTH, SAFETY AND ENVIRONMENTAL PROTECTION COMMITTEE May 10, 2011 Richland, WA Topics in this...

439

Preservation of FFTF Data Related to Passive Safety Testing  

SciTech Connect

One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980s. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: 1) to verify natural circulation as a reliable means to safely remove decay heat, 2) to extend passive safety experience to a large-size LMR and obtain data for validating design analysis computer codes, and 3) to develop and test passive safety enhancements that might be used for future LMRs. These tests were designed to provide data sufficient to allow separation of fuel temperature effects from structural temperature effects. The data developed through this testing program were used to verify the predictive capability of passive safety analysis methods as well as provide a data base for calibrating design tools such as the SASSYS/SAS4A codes. These tests were instrumental in improving understanding of reactivity feedback mechanisms in LMRs and demonstrating passive safety margins available in an LMR. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. This information may be of potential use for international exchanges with other LMR programs around the world. This information provides the basis for creating benchmarks for validating and testing large scale computer programs. All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. The test results information exists in several different formats depending upon the final stage of the test evaluation. Over 100 documents relevant to passive safety testing have been identified and are being recovered, scanned, and catalogued. Attempts to recover plant data tapes are also in progress. Documents related to passive safety testing are now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format compatible with a general search engine being developed by INL. The data from the FFTF passive safety tests provides experimental verification of structural reactivity effects that should be very useful to innovative designers seeking to optimize passive safety in the design of new LMRs.

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-10-01T23:59:59.000Z

440

Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants  

SciTech Connect

Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

Chang, T.Y.

1991-09-01T23:59:59.000Z

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.


441

Safety, Security & Fire Report  

E-Print Network (OSTI)

2013 Safety, Security & Fire Report Stanford University #12;Table of Contents Public Safety About the Stanford University Department of Public Safety Community Outreach & Education Programs Emergency Access Transportation Safety Bicycle Safety The Jeanne Clery and Higher Education Act Timely Warning

Straight, Aaron

442

Technical Review Report for the Model 9978-96 Package Safety Analysis Report for Packaging (S-SARP-G-00002, Revision 1, March 2009)  

SciTech Connect

This Technical Review Report (TRR) documents the review, performed by Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the Department of Energy (DOE), on the 'Safety Analysis Report for Packaging (SARP), Model 9978 B(M)F-96', Revision 1, March 2009 (S-SARP-G-00002). The Model 9978 Package complies with 10 CFR 71, and with 'Regulations for the Safe Transport of Radioactive Material-1996 Edition (As Amended, 2000)-Safety Requirements', International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9978 Packaging is designed, analyzed, fabricated, and tested in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC). The review presented in this TRR was performed using the methods outlined in Revision 3 of the DOE's 'Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages'. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's Regulatory Guide 7.9, i.e., 'Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material'. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9978 Packaging is a single containment package, using a 5-inch containment vessel (5CV). It uses a nominal 35-gallon drum package design. In comparison, the Model 9977 Packaging uses a 6-inch containment vessel (6CV). The Model 9977 and Model 9978 Packagings were developed concurrently, and they were referred to as the General Purpose Fissile Material Package, Version 1 (GPFP). Both packagings use General Plastics FR-3716 polyurethane foam as insulation and as impact limiters. The 5CV is used as the Primary Containment Vessel (PCV) in the Model 9975-96 Packaging. The Model 9975-96 Packaging also has the 6CV as its Secondary Containment Vessel (SCV). In comparison, the Model 9975 Packagings use Celotex{trademark} for insulation and as impact limiters. To provide a historical perspective, it is noted that the Model 9975-96 Packaging is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then-newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The Model 9978 Package has six Content Envelopes: C.1 ({sup 238}Pu Heat Sources), C.2 ( Pu/U Metals), C.3 (Pu/U Oxides, Reserved), C.4 (U Metal or Alloy), C.5 (U Compounds), and C.6 (Samples and Sources). Per 10 CFR 71.59 (Code of Federal Regulations), the value of N is 50 for the Model 9978 Package leading to a Criticality Safety Index (CSI) of 1.0. The Transport Index (TI), based on dose rate, is calculated to be a maximum of 4.1.

West, M

2009-03-06T23:59:59.000Z

443

Spent nuclear fuel project cold vacuum drying facility safety equipment list  

SciTech Connect

This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved.

IRWIN, J.J.

1999-02-24T23:59:59.000Z

444

Reactor operation safety information document  

SciTech Connect

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Not Available

1990-01-01T23:59:59.000Z

445

Final Notice of Violation  

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

8,2011 8,2011 CERTIFIED MAIL RETURN RECEIPT REQUESTED Mr. Jolm J. Grossenbacher Director, Idaho National Laboratory and President, Battelle Energy Alliance, LLC P. O. Box 1625 Idaho Falls, Idaho 83415-3695 SEA-2011-01 Dear Mr. Grossenbacher: Pursuant to section 234B of the Atomic Energy Act of 1954, as amended, (the Act), and the Department of Energy's (DOE) regulations at 10 C.F.R . §§ 824.4(a)(3) and 824.7(b), DOE is issuing this Final Notice of Violation (FNOV) to Battelle Energy Alliance, LLC (BEA) for multiple violations of classified information security requirements. The FNOV is based upon the Office of Health, Safety and Security's Office of Enforcement May 11, 2010, Investigation Report and an evaluation of the evidence presented to DOE by BEA, including BEA's final

446

Final Reminder:  

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

Final Reminder: Final Reminder: Final Reminder: Please save your $SCRATCH and $SCRATCH2 imporant files by 4/30/12 April 27, 2012 by Helen He (0 Comments) Franklin batch system is drained, and all batch queues are stopped as of 4/26 23:59pm. This is the final reminder that please make sure to save important files on your Franklin $SCRATCH and $SCRATCH2. ALL FILES THERE WILL BE DELETED, and there will be no mechanisms to recover any of the files after May 1. Mon Apr 30: Last day to retrieve files from Franklin scratch file systems Mon Apr 30, 23:59: User logins are disabled If you need help or have any concerns, please contact "consult at nersc dot gov". Post your comment You cannot post comments until you have logged in. Login Here. Comments No one has commented on this page yet.

447

Final Report  

SciTech Connect

This the final report for the project "Large-Scale Optimization for Bayesian Inference in Complex Systems," for the work in the group of the co-PI George Biros.

Biros, George

2014-08-18T23:59:59.000Z

448

Final Report  

SciTech Connect

This document constitutes the Final Report for award DE-FC02-06ER41446 as required by the Office of Science. It summarizes accomplishments and provides copies of scientific publications with significant contribution from this award.

DeTar, Carleton [P.I.

2012-12-10T23:59:59.000Z

449

ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network (OSTI)

ENVIRONMENTAL HEALTH & SAFETY ORIENTATION HANDBOOK Environmental Health and Safety Office safety & Safety Office 494-2495 (Phone) 494-2996 (Fax) Safety.Office@dal.ca (E-mail) www.dal.ca/safety (Web) Radiation Safety Office 494-1938 (Phone) 494-2996 (Fax) Melissa.Michaud@dal.ca (E-mail) University

Brownstone, Rob

450

Electrical Safety Occurrences | Department of Energy  

Office of Environmental Management (EM)

- April 2013 An analysis of the Occurrence Reporting and Processing System (ORPS) reports that was requested by the Electrical Safety Community for information exchange and...

451

Predictive Value of Phase I Trials for Safety in Later Trials and Final Approved Dose: Analysis of 61 Approved Cancer Drugs  

Science Journals Connector (OSTI)

...2012 were identified on the FDA website (7). Agents approved for...leading to approval. An extensive search was concomitantly done through...TIBCO Software Inc). Results Search results Between January 1990...BH, Squitieri L, Stallings SC, Halpern EF, Chabner BA...

Denis L. Jardim; Kenneth R. Hess; Patricia LoRusso; Razelle Kurzrock; and David S. Hong

2014-01-15T23:59:59.000Z

452

Asymptotic Safety  

E-Print Network (OSTI)

Asymptotic safety is a set of conditions, based on the existence of a nontrivial fixed point for the renormalization group flow, which would make a quantum field theory consistent up to arbitrarily high energies. After introducing the basic ideas of this approach, I review the present evidence in favor of an asymptotically safe quantum field theory of gravity.

R. Percacci

2008-11-18T23:59:59.000Z

453

Gas Pipeline Safety (Indiana)  

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

This section establishes the Pipeline Safety Division within the Utility Regulatory Commission to administer federal pipeline safety standards and establish minimum state safety standards for...

454

Exploration of High-dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization: A User's Guide to TopoXG*  

SciTech Connect

Large-scale simulation datasets can be modeled as high-dimensional scalar functions defined over a discrete sample of the domain. The goals of our proposed research are two-fold. First, we would like to provide structural analysis of 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. TopoXG is a software package that is designed to address these goals. The unique contribution of TopoXG lies in 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 TopoXG, by highlighting its analysis and visualization capabilities, and giving several use cases involving datasets from nuclear reactor safety simulations.

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

2012-10-01T23:59:59.000Z

455

Criticality Safety Analysis on the Mixed Be, Nat-U, and C (Graphite) Reflectors in 55-Gallon Waste Drums and Their Equivalents for HWM Applications  

SciTech Connect

The objective of this analysis is to develop and establish the technical basis on the criticality safety controls for the storage of mixed beryllium (Be), natural uranium (Nat-U), and carbon (C)/graphite reflectors in 55-gallon waste containers and/or their equivalents in Hazardous Waste Management (HWM) facilities. Based on the criticality safety limits and controls outlined in Section 3.0, the operations involving the use of mixed-reflector drums satisfy the double-contingency principle as required by DOE Order 420.1 and are therefore criticality safe. The mixed-reflector mass limit is 120 grams for each 55-gallon drum or its equivalent. a reflector waiver of 50 grams is allowed for Be, Nat-U, or C/graphite combined. The waived reflectors may be excluded from the reflector mass calculations when determining if a drum is compliant. The mixed-reflector drums are allowed to mix with the typical 55-gallon one-reflector drums with a Pu mass limit of 120 grams. The fissile mass limit for the mixed-reflector container is 65 grams of Pu equivalent each. The corresponding reflector mass limits are 300 grams of Be, and/or 100 kilograms of Nat-U, and/or 110 kilograms of C/graphite for each container. All other unaffected control parameters for the one-reflector containers remain in effect for the mixed-reflector drums. For instance, Superior moderators, such as TrimSol, Superla white mineral oil No. 9, paraffin, and polyethylene, are allowed in unlimited quantities. Hydrogenous materials with a hydrogen density greater than 0.133 gram/cc are not allowed. Also, an isolation separation of no less than 76.2 cm (30-inch) is required between a mixed array and any other array. Waste containers in the action of being transported are exempted from this 76.2-cm (30-inch) separation requirement. All deviations from the CS controls and mass limits listed in Section 3.0 will require individual criticality safety analyses on a case-by-case basis for each of them to confirm their criticality safety prior to their deployment and implementation.

Chou, P

2011-12-14T23:59:59.000Z

456

ENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION  

E-Print Network (OSTI)

ENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION SIMON FRASER UNIVERSITY ENVIRONMENTAL HEALTH & SAFETY DEPARTMENT Discovery Park - MTF 8888 University Drive Burnaby, British Columbia Canada V5 FOOTWEAR 23867 TRADES & CONSTRUCTION 23867 TRANSPORT OF DANGEROUS GOODS 27265 WORKPLACE ENVIRONMENT 23867

457

DOE/EIS-0287-SA-01: Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (June 2005)  

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

7 7 -SA-Ol SUPPLEMENT ANALYSIS For The Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement June 2005 United States Department of Energy Idaho Operations Office 1.0 2.0 3.0 4.0 5.0 6.0 DOEÆIS-0287 -SA-O 1 TABLE OF CONTENTS Introduction......................................................................................................................... 4 Background......................................................................................................................... 4 Areas of Review.................................................................................................................. 6 3.1 3.2 3.3 3.4 Proposed Waste Treatment Technology.......... .......................................................

458

EVALUATING THE EFFECTIVENESS OF THE SAFETY INVESTMENT PROGRAM  

E-Print Network (OSTI)

EVALUATING THE EFFECTIVENESS OF THE SAFETY INVESTMENT PROGRAM (SIP) POLICIES FOR OREGON Final Report SPR 651 #12;#12;EVALUATING THE EFFECTIVENESS OF THE SAFETY INVESTMENT PROGRAM (SIP) POLICIES the Effectiveness of the Safety Investment Program (SIP) Policies for Oregon 5. Report Date October 2009 6

Bertini, Robert L.

459

Machine Shop Safety Tips & Safety Guidelines GENERAL SAFETY TIPS  

E-Print Network (OSTI)

Machine Shop Safety Tips & Safety Guidelines GENERAL SAFETY TIPS · Safety glasses with side shields distance away from moving machine parts, work pieces, and cutters. · Use hand tools for their designed to oil, clean, adjust, or repair any machine while it is running. Stop the machine and lock the power

Veiga, Pedro Manuel Barbosa

460

Final Report  

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

Final Final Report to Improved Reservoir Access Through Refracture Treatments in Tight Gas Sands and Gas Shales 07122-41.FINAL June 2013 PI Mukul M. Sharma The University of Texas at Austin 200 E. Dean Keeton St. Stop C0300 Austin, Texas 78712 (512) 471---3257 msharma@mail.utexas.edu LEGAL NOTICE This report was prepared by The University of Texas at Austin as an account of work sponsored by the Research Partnership to Secure Energy for America, RPSEA. Neither RPSEA members of RPSEA, the National Energy Technology Laboratory, the U.S. Department of Energy, nor any person acting on behalf of any of the entities: a. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED WITH RESPECT TO ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS DOCUMENT, OR THAT THE

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.


461

CRAD, Facility Safety - Unreviewed Safety Question Requirements...  

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

a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Facility Safety - Unreviewed Safety Question Requirements...

462

Safety Shoe Mobile  

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

Safety and Training Safety Notices Safety Shoe Mobile The Safety Shoe Mobile comes to Argonne every Monday on the following schedule: 200 Area: 0800 - 1200 360 Area: 1300 - 1630...

463

OCCUPATIONAL SAFETY and HEALTH  

E-Print Network (OSTI)

MARYLAND OCCUPATIONAL SAFETY and HEALTH ACT safety and health protection on the job STATE OCCUPATIONAL SAFETY AND HEALTH STANDARDS, AND OTHER APPLICABLE REGULATIONS MAY BE OBTAINED FROM Complaints about State Program administration may be made to Regional Administrator, Occupational Safety

Weaver, Harold A. "Hal"

464

OCCUPATIONAL HEALTH AND SAFETY  

E-Print Network (OSTI)

OCCUPATIONAL HEALTH AND SAFETY MANAGEMENT SYSTEM Department of Occupational Health and Safety Revised December 2009 #12;Occupational Health and Safety (OHS) Management System 1. Introduction.............................................................................................................. 3 2.2 Management of Health and Safety

465

Electrical Safety  

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

NOT MEASUREMENT NOT MEASUREMENT SENSITIVE DOE HANDBOOK ELECTRICAL SAFETY DOE-HDBK-1092-2013 July 2013 Superseding DOE-HDBK-1092-2004 December 2004 U.S. Department of Energy AREA SAFT Washington, D.C.20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-HDBK-1092-2013 Available on the Department of Energy Technical Standards Program Web site at http://www.hss.doe.gov/nuclearsafety/techstds/ ii DOE-HDBK-1092-2013 FOREWORD 1. This Department of Energy (DOE) Handbook is approved for use by the Office of Health, Safety and Security and is available to all DOE components and their contractors. 2. Specific comments (recommendations, additions, deletions, and any pertinent data) to enhance this document should be sent to: Patrick Tran

466

Safety Notices  

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

Safety Notices Safety Notices Fatigue August 2011 Sleep deprivation and the resulting fatigue can adversely affect manual dexteri- ty, reaction time, alertness, and judgment, resulting in people putting themselves and their co-workers at risk. Liquid-Gas Cylinder Handtruck Awareness May 2011 Failure of a spring assembly can result in a loss of control, allowing the Dewar to become separated from the hand truck, leading to a very dangerous situation. Safe Transport of Hazardous Materials February 2011 APS users are reminded that hazardous materials, including samples, cannot be packed in personal luggage and brought on public transport. Electrical Incidents September 2010 Two minor electrical incidents in the past months at the APS resulted in a minor shock from inadequately grounded equipment, and a damaged stainless

467

Safety harness  

DOE Patents (OSTI)

A safety harness to be worn by a worker, especially a worker wearing a plastic suit thereunder for protection in a radioactive or chemically hostile environment, which safety harness comprises a torso surrounding portion with at least one horizontal strap for adjustably securing the harness about the torso, two vertical shoulder straps with rings just forward of the of the peak of the shoulders for attaching a life-line and a pair of adjustable leg supporting straps releasibly attachable to the torso surrounding portion. In the event of a fall, the weight of the worker, when his fall is broken and he is suspended from the rings with his body angled slightly back and chest up, will be borne by the portion of the leg straps behind his buttocks rather than between his legs. Furthermore, the supporting straps do not restrict the air supplied through hoses into his suit when so suspended.

Gunter, Larry W. (615 Sand Pit Rd., Leesville, SC 29070)

1993-01-01T23:59:59.000Z

468

Explosives Safety  

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

212-2012 212-2012 June 2012 DOE STANDARD EXPLOSIVES SAFETY U.S. Department of Energy AREA SAFT Washington, DC 20585 MEASUREMENT SENSITIVE DOE-STD-1212-2012 i TABLE OF CONTENTS CHAPTER I. PURPOSE, SCOPE and APPLICABILITY, EXEMPTIONS, WAIVERS, ABBREVIATIONS, ACRONYMS, AND DEFINITIONS .......... 1 1.0. PURPOSE ............................................................................................................. 1 1.1. Scope and Applicability.............................................................................. 1 2.0. STANDARD ADMINISTRATION AND MANAGEMENT ...................................... 3 3.0. EXEMPTIONS ....................................................................................................... 4

469

Safety valve  

DOE Patents (OSTI)

The safety valve contains a resilient gland to be held between a valve seat and a valve member and is secured to the valve member by a sleeve surrounding the end of the valve member adjacent to the valve seat. The sleeve is movable relative to the valve member through a limited axial distance and a gap exists between said valve member and said sleeve.

Bergman, Ulf C. (Malmoe, SE)

1984-01-01T23:59:59.000Z

470

Setting clear expectations for safety basis development  

SciTech Connect

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

MORENO, M.R.

2003-05-03T23:59:59.000Z

471

Supplement Analysis for the Final Environmental Impact Statement for the Continued Operation of the Pantex Plant and Associated Storage of Nuclear Weapon Components  

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

D D E P A R T M E N T O F E N E R G Y U N I T E D S T A T E S O F A M E R I C A SUPPLEMENT ANALYSIS FOR THE FINAL ENVIRONMENTAL IMPACT STATEMENT FOR THE CONTINUED OPERATION OF THE PANTEX PLANT AND ASSOCIATED STORAGE OF NUCLEAR WEAPON COMPONENTS DOE/EIS-0225/SA-03 United States Department of Energy National Nuclear Security Administration Pantex Site Operations P.O. Box 30030 Amarillo, Texas 79120-0030 February 2003 i Summary The U.S. Department of Energy's (DOE's) National Environmental Policy Act (NEPA) Implementing Procedures at 10 CFR 1021.330(d) require evaluation of its site-wide environmental impact statements (EISs) at least every 5 years by preparation of a supplement analysis (SA), as provided in 10 CFR 1021.314. Based on the SA, a determination is made as to whether the existing EIS remains

472

Operations research and systems analysis of geopressured-geothermal energy in Louisiana. Final report for the period June 1, 1978-August 31, 1979  

SciTech Connect

The primary purpose was to provide a projection of the probable future contribution of the geopressured-geothermal energy resource in Louisiana to the overall energy requirements of the nation. A number of associated objectives were emphasized: namely, development of the tools and methodology for performing economic analyses, application of these tools to specific prospects about which adequate resource assessments have been made, identification of the impediments to resource development, and socio-economic analysis of the impact of development of the resource on these specific prospects. An overview of the geopressured-geothermal resource activities in Louisiana is provided first, followed by a detailed discussion and review of the achievements of this project. Finally the major conclusions and findings of this project with respect to commercial viability, impediments, and social and economic impact are presented, and recommendations are made for future systems analysis work.

Johnson, A.E. Jr.

1980-11-01T23:59:59.000Z

473

Models for residential- and commercial-sector energy-conservation analysis: applications, limitations, and future potential. Final report  

SciTech Connect

This report reviews four of the major models used by the Department of Energy (DOE) for energy conservation analyses in the residential- and commercial-building sectors. The objective is to provide a critical analysis of how these models can serve as tools for DOE and its Conservation Policy Office in evaluating and quantifying their policy and program requirements. For this, the study brings together information on the models' analytical structure and their strengths and limitations in policy applications these are then employed to assess the most-effective role for each model in addressing future issues of buildings energy-conservation policy and analysis. The four models covered are: Oak Ridge Residential Energy Model; Micro Analysis of Transfers to Households/Comprehensive Human Resources Data System (MATH/CHRDS) Model; Oak Ridge Commercial Energy Model; and Brookhaven Buildings Energy Conservation Optimization Model (BECOM).

Cole, Henry E.; Fullen, Robert E.

1980-09-01T23:59:59.000Z

474

A Complexity Science-Based Framework for Global Joint Operations Analysis to Support Force Projection: LDRD Final Report.  

SciTech Connect

The military is undergoing a significant transformation as it modernizes for the information age and adapts to address an emerging asymmetric threat beyond traditional cold war era adversaries. Techniques such as traditional large-scale, joint services war gaming analysis are no longer adequate to support program evaluation activities and mission planning analysis at the enterprise level because the operating environment is evolving too quickly. New analytical capabilities are necessary to address modernization of the Department of Defense (DoD) enterprise. This presents significant opportunity to Sandia in supporting the nation at this transformational enterprise scale. Although Sandia has significant experience with engineering system of systems (SoS) and Complex Adaptive System of Systems (CASoS), significant fundamental research is required to develop modeling, simulation and analysis capabilities at the enterprise scale. This report documents an enterprise modeling framework which will enable senior level decision makers to better understand their enterprise and required future investments.

Lawton, Craig R.

2015-01-01T23:59:59.000Z

475

The Office of Health, Safety and Security  

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

HSS Logo Department of Energy Seal HSS Logo Department of Energy Seal Left Tab SEARCH Right Tab TOOLS Right Tab Left Tab HOME Right Tab Left Tab ABOUT US Right Tab Left Tab FUNCTIONS Right Tab Left Tab RESOURCES Right Tab Left Tab NEWSFEEDS Right Tab Left Tab VIDEOS Right Tab Left Tab EVENTS Environmental Protection, Sustainability Support & Corporate Safety Analysis HS-20 Home Mission & Functions » Office of Nuclear Safety, Quality Assurance & Environment » Sustainability Support » Environmental Policy & Assistance » Corporate Safety Programs » Analysis Program Contacts What's New? Sustainability Support Environment Corporate Safety Programs Analysis Environment Environmental Policy Environmental Guidance Environmental Reports Environmental Management System

476

A Study on Uncertainty Analysis of Safety Systems of Advanced Heavy Water Reactor using Fuzzy Set Theory  

Science Journals Connector (OSTI)

Inherent to any reliability calculation is a degree of uncertainty in ... ) are also being used in the risk analysis for quantifying the basic event uncertainty and ... of probabilistic and fuzzy methodologies fo...

Rao K. Durga; V. Gopika; M. H. Prasad

2004-01-01T23:59:59.000Z

477

Russell Furr Laboratory Safety &  

E-Print Network (OSTI)

Russell Furr Director 8/20/13 Laboratory Safety & Compliance #12;#12;Research Safety Full Time Students Part- Time #12; Organizational Changes Office of Research Safety Research Safety Advisors Safety Culture Survey Fire Marshal Inspections Laboratory Plans Review New Research Safety Initiatives

478

FAQS Qualification Card - Criticality Safety | Department of Energy  

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

Criticality Safety Criticality Safety FAQS Qualification Card - Criticality Safety A key element for the Department's Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA). For each functional area, the FAQS identify the minimum technical competencies and supporting knowledge and skills for a typical qualified individual working in the area. FAQC-CriticalitySafety.docx Description Criticality Safety Qualification Card More Documents & Publications FAQS Gap Analysis Qualification Card - Criticality Safety

479

Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Observation of Waste Treatment and Immobilization Plant LAW Melter and Melter Off-gas Process System Hazards Analysis _Oct 21-31  

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

HSS Independent Activity Report - HSS Independent Activity Report - Rev. 0 Report Number: HIAR-WTP-2013-10-21 Site: Hanford Site Subject: Office of Enforcement and Oversight's Office of Safety and Emergency Management Evaluations Activity Report for Observation of Waste Treatment and Immobilization Plant Low Activity Waste Melter and Melter Off-gas Process System Hazards Analysis Activities Dates of Activity : 10/21/13 - 10/31/13 Report Preparer: James O. Low Activity Description/Purpose: The Office of Health, Safety and Security (HSS), Office of Safety and Emergency Management Evaluations (Independent Oversight) reviewed the Insight software hazard evaluation (HE) tables for hazard analysis (HA) generated to date for the Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) Melter and Off-gas systems, observed a

480

SOLERAS - Solar Controlled Environment Agriculture Project. Final report, Volume 2. Battelle Columbus Laboratories system requirements definition and system analysis  

SciTech Connect

System specifications, design criteria, and representative weather data necessary for the system evolution and preliminary design are generated. A detailed analysis and evaluation of the commercial-sized controlled environment agriculture system coupled with a solar energy system was conducted. A simulation model to test the performance of the greenhouse is presented. (BCS)

Not Available

1981-01-01T23:59:59.000Z

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481

Reliability of steam-turbine rotors. Task 1. Lifetime prediction analysis system. Final report. [Using STRAP and SAFER computer codes and boresonic data  

SciTech Connect

Task 1 of RP 502, Reliability of Steam Turbine Rotors, resulted in the development of a computerized lifetime prediction analysis system (STRAP) for the automatic evaluation of rotor integrity based upon the results of a boresonic examination of near-bore defects. Concurrently an advanced boresonic examination system (TREES), designed to acquire data automatically for lifetime analysis, was developed and delivered to the maintenance shop of a major utility. This system and a semi-automated, state-of-the-art system (BUCS) were evaluated on two retired rotors as part of the Task 2 effort. A modified nonproprietary version of STRAP, called SAFER, is now available for rotor lifetime prediction analysis. STRAP and SAFER share a common fracture analysis postprocessor for rapid evaluation of either conventional boresonic amplitude data or TREES cell data. The final version of this postprocessor contains general stress intensity correlations for elliptical cracks in a radial stress gradient and provision for elastic-plastic instability of the ligament between an imbedded crack and the bore surface. Both linear elastic and ligament rupture models were developed for rapid analysis of linkup within three-dimensional clusters of defects. Bore stress-rupture criteria are included, but a creep-fatigue crack growth data base is not available. Physical and mechanical properties of air-melt 1CrMoV forgings are built into the program; however, only bounding values of fracture toughness versus temperature are available. Owing to the lack of data regarding the probability of flaw detection for the boresonic systems and of quantitative verification of the flaw linkup analysis, automatic evlauation of boresonic results is not recommended, and the lifetime prediction system is currently restricted to conservative, deterministic analysis of specified flaw geometries.

Nair, P.K.; Pennick, H.G.; Peters, J.E.; Wells, C.H.

1982-12-01T23:59:59.000Z

482

Safety | Data.gov  

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

Safety Safety Safety Data/Tools Apps Challenges Resources Blogs Let's Talk Safety Welcome to the Safety Community The Safety Community is where data and insight are combined to facilitate a discussion around and awareness of our Nation's public safety activities. Whether you are interested in crime, roadway safety, or safety in the workplace, we have something for you. Check out the data, browse and use the apps, and be part of the discussion. Check out talks from the White House Safety Datapalooza Previous Pause Next One year of public safety data at Safety.Data.gov! Safety NHTSA releases SaferCar APIs and mobile app NHTSA releases SaferCar APIs and mobile app View More Todd Park, U.S. Chief Technology Officer at the Safety Datapalooza View More New APIs New APIs FRA launches new safety data dashboard and APIs.

483

Modified fuzzy algorithm based safety analysis of nuclear energy for sustainable hydrogen production in climate change prevention  

Science Journals Connector (OSTI)

Abstract The high temperature coolant of nuclear power plants (NPPs) has been investigated for the hydrogen production, which could be a major role of a green energy promotion. An accident of the high temperature gas cooled reactor (HTGR) is modeled for the stabilized hydrogen production using nuclear energy. For the clean energy resource pursuit in preventing the climate change, the hydrogen is one of very attractive energy sources. The non-operational data could be produced by the fuzzy set theory which is one of non-linear complex algorithms. So, the result can show the possibility of the event happening instead of the exact solutions. The random numbers are generated for membership numbers of the fuzzy function. The event manipulation is done by new membership numbers for the propagations. The final result is 1.0 in 8 times during 100months. So, the frequency is 0.08, or 8% of successful long-term cooling by conduction.

Tae Ho Woo

2014-01-01T23:59:59.000Z

484

Environmental Health & Safety Office of Radiological Safety  

E-Print Network (OSTI)

Environmental Health & Safety Office of Radiological Safety Page 1 of 2 FORM LU-1 Revision 01 1 safety training and submit this registration to the LSO prior to use of Class 3B or 4 lasers. A copy will be returned to the Laser Supervisor to be filed in the Laboratory Laser Safety Notebook. Both the Laser

Houston, Paul L.

485

Environmental Health and Instructional Safety Employee Safety  

E-Print Network (OSTI)

Environmental Health and Instructional Safety #12;Employee Safety Page 1 To our University an environment for students, faculty, staff, and visitors that will not adversely affect their health and safety task that is unsafe or hazardous. Environmental Health and Instructional Safety can assist departments

de Lijser, Peter

486

Safety Share from National Safety Council  

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

Slide Presentation by Joe Yanek, Fluor Government Group. National Safety Council Safety Share. The Campbell Institute is the Environmental, Health and Safety (EHS) Center of Excellence at the National Safety Council and provides a Forum for Leaders in EHS to exchange ideas and collaborate across industry sectors and organizational types.