Powered by Deep Web Technologies
Note: This page contains sample records for the topic "documented 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

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

Office of Environmental Management (EM)

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

2

B PLANT DOCUMENTED SAFETY ANALYSIS  

SciTech Connect (OSTI)

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

3

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...

4

SNF fuel retrieval sub project safety analysis document  

SciTech Connect (OSTI)

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

5

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

SciTech Connect (OSTI)

This Safety Evaluation Report (SER) documents the Department of Energy’s (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

6

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

SciTech Connect (OSTI)

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

Alan Bond

2001-04-01T23:59:59.000Z

7

K West integrated water treatment system subproject safety analysis document  

SciTech Connect (OSTI)

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

SEMMENS, L.S.

1999-02-24T23:59:59.000Z

8

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

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

Technical Safety Requirements - December 2, 2014 (EA CRAD 31-07, Rev. 0) More Documents & Publications CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08...

9

Planning Document for an NBSR Conversion Safety Analysis Report  

SciTech Connect (OSTI)

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

10

Documented Safety Analysis for the B695 Segment  

SciTech Connect (OSTI)

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

Laycak, D

2008-09-11T23:59:59.000Z

11

Documented Safety Analysis for the Waste Storage Facilities  

SciTech Connect (OSTI)

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

Laycak, D

2008-06-16T23:59:59.000Z

12

Documented Safety Analysis for the Waste Storage Facilities March 2010  

SciTech Connect (OSTI)

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

Laycak, D T

2010-03-05T23:59:59.000Z

13

Acceptable NSLS Safety Documentation  

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

Acceptable NSLS Safety Documentation Print NSLS users who have completed NSLS Safety Module must present a copy of one of the following documents to receive ALS 1001: Safety at the...

14

Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade  

SciTech Connect (OSTI)

The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

Gregg L. Sharp; R. T. McCracken

2003-06-01T23:59:59.000Z

15

Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade  

SciTech Connect (OSTI)

The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

Sharp, G.L.; McCracken, R.T.

2003-05-13T23:59:59.000Z

16

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

SciTech Connect (OSTI)

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

17

324 Building safety basis criteria document  

SciTech Connect (OSTI)

The Safety Basis Criteria document describes the proposed format, content, and schedule for the preparation of an updated Safety Analysis Report (SAR) and Operational Safety Requirements document (OSR) for the 324 Building. These updated safety authorization basis documents are intended to cover stabilization and deactivation activities that will prepare the facility for turnover to the Environmental Restoration Contractor for final decommissioning. The purpose of this document is to establish the specific set of criteria needed for technical upgrades to the 324 Facility Safety Authorization Basis, as required by Project Hanford Procedure HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval.

STEFFEN, J.M.

1999-06-02T23:59:59.000Z

18

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

SciTech Connect (OSTI)

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

Garvin, L.J.

1997-04-28T23:59:59.000Z

19

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

SciTech Connect (OSTI)

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

Not Available

1988-05-01T23:59:59.000Z

20

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

SciTech Connect (OSTI)

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

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

Boyd D. Christensen

2009-05-01T23:59:59.000Z

22

Complete Experiment Safety Documentation  

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

Safety Sheet If you did not submit a General User Proposal, you must submit an ESS one month prior to arrival at the ALS. 2. Biological, Radioactive, and Hazardous...

23

Complete Experiment Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group and UserofProteinNewsatCompiler Compiling and

24

Complete Experiment Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group and UserofProteinNewsatCompiler Compiling andComplete

25

Complete Experiment Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group and UserofProteinNewsatCompiler Compiling

26

Reactor operation safety information document  

SciTech Connect (OSTI)

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

27

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

SciTech Connect (OSTI)

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

28

K Basin safety analysis  

SciTech Connect (OSTI)

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

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

1994-12-16T23:59:59.000Z

29

Spent nuclear fuel project criteria document -- Cold Vacuum Drying Facility phase 2 safety analysis report  

SciTech Connect (OSTI)

The criteria document provides the criteria and guidance for developing the SNF CVDF Phase 2 SAR. This SAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, and installation acceptance testing of the CVDF systems.

Garvin, L.J.

1998-07-03T23:59:59.000Z

30

DISCIPLINE OF RADIATION THERAPY SAFETY DOCUMENT  

E-Print Network [OSTI]

DISCIPLINE OF RADIATION THERAPY SAFETY DOCUMENT Contents Page 1. Health & Safety arrangements 1 Inspection 1.4 Hazard Reporting 1.5 Security in College 1.6 Out of Hours Working 1.7 Management of Work.7 Management of Work-related Stress Those experiencing symptoms of work-related stress should raise this matter

O'Mahony, Donal E.

31

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

SciTech Connect (OSTI)

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

Not Available

1989-01-25T23:59:59.000Z

32

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation  

SciTech Connect (OSTI)

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

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

2000-03-23T23:59:59.000Z

33

PUREX Deactivation Health and Safety documentation  

SciTech Connect (OSTI)

The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D&D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety.

Dodd, E.N. III

1995-01-01T23:59:59.000Z

34

PORTAL Aggregation Analysis and Documentation  

E-Print Network [OSTI]

PORTAL Aggregation Analysis and Documentation Kristin A. Tufte Portland State University Introduction This document describes the aggregation procedures for PORTAL . Analysis of different aggregation system. Details on the status values and their descriptions can be found in the SWARM documentation

Bertini, Robert L.

35

Documentation Integrity for Safety-Critical Applications: The COHERE Project  

E-Print Network [OSTI]

. Keywords Authoring interface, documentation integrity, consistency 1. INTRODUCTION This paper reportsDocumentation Integrity for Safety-Critical Applications: The COHERE Project David G. Novick-critical systems. Following a set of documentation integrity maxims, the project developed two generations

Novick, David G.

36

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

Broader source: Energy.gov [DOE]

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

37

GARS Directorate Environment, Safety, & Health Records/Documents Management  

E-Print Network [OSTI]

GARS Directorate Environment, Safety, & Health Records/Documents Management Rev. 6 4/24/13 Topic documentation for Environmental Management System (EMS) Program CONTROLLED DOCUMENT ENV-1.K.1 - Cut off annually. Occupational Health & Safety (OHSAS 18001) Management System Description (Manual) The written OSH program

Ohta, Shigemi

38

Hanford safety analysis and risk assessment handbook (SARAH)  

SciTech Connect (OSTI)

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

39

HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)  

SciTech Connect (OSTI)

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

EVANS, C B

2004-12-21T23:59:59.000Z

40

National Spherical Torus Experiment-Upgrade SAFETY ASSESSMENT DOCUMENT  

E-Print Network [OSTI]

-00 Revision 6 DECEMBER 2014 Prepared by: _______________________ Jerry Levine Head, PPPL ESH&S Reviewed by Chair, PPPL ES&H Executive Board #12;#12;NSTX-U SAFETY ASSESSMENT DOCUMENT (SAD) Table of Contents 1

Princeton Plasma Physics Laboratory

Note: This page contains sample records for the topic "documented 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

Lawrence Berkeley National Laboratory Safety Assessment Document (SAD)  

E-Print Network [OSTI]

Lawrence Berkeley National Laboratory Safety Assessment Document (SAD) for the Advanced Light Assessment Document, Rev. 7 (May 29, 2009) ii Signature Page for Rev. 7 of the ALS SAD Prepared by: ALS EHS Program Manager Date: Reviewed by: ALS Deputy Division Director Date: ALS Deputy for Operations

Knowles, David William

42

Madonne: Document Image Analysis Techniques for Cultural Heritage Documents  

E-Print Network [OSTI]

Madonne: Document Image Analysis Techniques for Cultural Heritage Documents Jean-Marc Ogier and Karl Tombre Abstract. This paper presents the Madonne project, a French initiative to use document image anal- ysis techniques for the purpose of preserving and exploiting heritage documents. 1

Boyer, Edmond

43

Nuclear Criticality Safety Application Guide: Safety Analysis Report Update Program  

SciTech Connect (OSTI)

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

Not Available

1994-02-01T23:59:59.000Z

44

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

SciTech Connect (OSTI)

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

Beaulieu, R. A.

2013-10-03T23:59:59.000Z

45

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

SciTech Connect (OSTI)

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

NONE

1995-01-01T23:59:59.000Z

46

SYNTHESIS OF SAFETY ANALYSIS AND FIRE HAZARD ANALYSIS METHODOLOGIES  

SciTech Connect (OSTI)

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

Coutts, D

2007-04-17T23:59:59.000Z

47

Hanford Site Wide Transportation Safety Document [SEC 1 Thru 3  

SciTech Connect (OSTI)

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

MCCALL, D L

2002-06-01T23:59:59.000Z

48

Events Beyond Design Safety Basis Analysis | Department of Energy  

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

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...

49

Health and Safety Policy This Health and Safety Policy document comprises a Policy Statement and details the  

E-Print Network [OSTI]

) The Corporate Manslaughter and Corporate Homicide Act and (ii) The Health & Safety at Work etc. Act. UnderHealth and Safety Policy UHSP/0/10 This Health and Safety Policy document comprises a Policy for implementing and monitoring them (UEP/0/10). The University has a Food Safety Policy Statement that covers all

Birmingham, University of

50

SAFEGUARDS AND SECURITY INTEGRATION WITH SAFETY ANALYSIS  

SciTech Connect (OSTI)

The objective of this paper is to share the Savannah River Site lessons learned on Safeguards and Security (S&S) program integration with K-Area Complex (KAC) safety basis. The KAC Documented Safety Analysis (DSA), is managed by the Washington Savannah River Company (WSRC), and the S&S program, managed by Wackenhut Services, Incorporated--Savannah River Site (WSI-SRS). WSRC and WSI-SRS developed a contractual arrangement to recognize WSI-SRS requirements in the KAC safety analysis. Design Basis Threat 2003 (DBT03) security upgrades required physical modifications and operational changes which included the availability of weapons which could potentially impact the facility safety analysis. The KAC DSA did not previously require explicit linkage to the S&S program to satisfy the safety analysis. WSI-SRS have contractual requirements with the Department of Energy (DOE) which are separate from WSRC contract requirements. The lessons learned will include a discussion on planning, analysis, approval of the controls and implementation issues.

Hearn, J; James Lightner, J

2007-04-13T23:59:59.000Z

51

Corporate Analysis of DOE Safety Performance  

Broader source: Energy.gov [DOE]

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

52

Document Image Analysis State of the Art  

E-Print Network [OSTI]

Document Image Analysis State of the Art and Technology Roadmap Eric Saund Area Manager, Perceptual Document Analysis Intelligent Systems Laboratory Palo Alto Research Center March, 2008 #12;Manfred Winslow rented a car... Every Document Tells a Story... Ervin Leasing billed a customer... Hugo Yepez hosted some

Saund, Eric

53

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

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

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

54

Patterns and Reuse in Document Engineering Keywords: Document Engineering, Document Analysis, Data Analysis, Data Modeling, Patterns, Reuse,  

E-Print Network [OSTI]

to business processes via Web-based services. The essence of Document Engineering is the analysis and designPatterns and Reuse in Document Engineering Keywords: Document Engineering, Document Analysis, Data and an Engineering Fellow at Commerce One. He founded or co-founded three companies, the last of which was Veo

Glushko, Robert J.

55

Safety analysis report for the Waste Storage Facility. Revision 2  

SciTech Connect (OSTI)

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

Bengston, S.J.

1994-05-01T23:59:59.000Z

56

Waste Isolation Pilot Plant Safety Analysis Report  

SciTech Connect (OSTI)

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

NONE

1995-11-01T23:59:59.000Z

57

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

SciTech Connect (OSTI)

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

58

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

Broader source: Energy.gov [DOE]

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

59

ASD Facility Hazard Analysis Document  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc Documentation RUCProductstwrmrAre the Effects of GlobalASCR User Facilities User

60

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

SciTech Connect (OSTI)

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

Not Available

1994-06-20T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

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

1993-04-15T23:59:59.000Z

62

PROPANE -C3H8 MSDS (Document # 001045) PAGE 1 OF 8 MATERIAL SAFETY DATA SHEET  

E-Print Network [OSTI]

PROPANE - C3H8 MSDS (Document # 001045) PAGE 1 OF 8 MATERIAL SAFETY DATA SHEET Prepared to U in an emergency? 1. PRODUCT IDENTIFICATION CHEMICAL NAME; CLASS: PROPANE - C3H8 Document Number: 001045 PRODUCT IN AIR ACGIH OSHA TLV STEL PEL STEL IDLH OTHER ppm ppm ppm ppm ppm Propane 74-98-6 > 96.0 Simple

Choi, Kyu Yong

63

CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energyon ArmedWaste andAccessCO2 Injection Begins in Illinois CO2Software Used in the2007 |-

64

CRAD, Facility Safety - Documented Safety Analysis | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTie Ltd:June 2015 <Ones |Laboratory, JuneDid y ouRev. 0)

65

CRAD, New Nuclear Facility Documented Safety Analysis and Technical Safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTie Ltd:June 2015 <Ones |Laboratory, JuneDid yRequirements - December

66

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

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

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

2001-10-24T23:59:59.000Z

67

Hot Cell Facility (HCF) Safety Analysis Report  

SciTech Connect (OSTI)

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

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

2000-11-01T23:59:59.000Z

68

Worker Safety and Health Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up from theDepartment of Dept.|Sindhu Jagadamma Women @Telecom (DraftWorker Safety and

69

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

SciTech Connect (OSTI)

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

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

1995-03-01T23:59:59.000Z

70

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

SciTech Connect (OSTI)

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

71

Safety analysis report for packaging (onsite) multicanister overpack cask  

SciTech Connect (OSTI)

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

72

Fuel Storage Facility Final Safety Analysis Report. Revision 1  

SciTech Connect (OSTI)

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

73

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

SciTech Connect (OSTI)

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

MCCOY, J.C.

1999-03-16T23:59:59.000Z

74

Safety assessment document for the Dynamic Test Complex B854  

SciTech Connect (OSTI)

A safety assessment was performed to determine if potential accidents at the 854 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (HE) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire. Seismic and extreme wind (including missiles) analyses indicate that the buildings are basically sound. The lightning protection system is in the process of being upgraded to meet AMCR 385-100. These buildings are located high above the dry creek bed so that a flood is improbable. The probability of high explosive detonation involving plutonium is very remote since the radioactive materials are encased and plutonium and HE are not permitted concurrently in the same area at Site 300. (The exception to this policy is that explosive actuating devices are sometimes located in assemblies containing fissile materials. However, an accidental actuation will not affect the safe containment of the plutonium within the assembly.) There is a remote possibility of an HE explosion involving uranium and beryllium since these are permitted in the same area.The possibility of a criticality accident is very remote since the fissile materials are doubly encased in stout metal containers. All operations involving these materials are independently reviewed and inspected by the Criticality Safety Office. It was determined that a fire was unlikely due to the low fire loading and the absence of ignition sources. It was also determined that the consequences of any accidents were reduced by the remote location of these facilities, their design, and by administrative controls.

Odell, B.N.; Pfeifer, H.E.

1981-12-11T23:59:59.000Z

75

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

Office of Environmental Management (EM)

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

76

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

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

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 21711 Comparison of Integrated Safety Analysis (ISA) and...

77

Satellite System Safety Analysis Using STPA  

E-Print Network [OSTI]

Traditional hazard analysis techniques based on failure models of accident causality, such as the probabilistic risk assessment (PRA) method currently used at NASA, are inadequate for analyzing safety at the system level. ...

Dunn, Nicholas Connor

2013-01-01T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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

NONE

1995-09-01T23:59:59.000Z

79

Autoclave nuclear criticality safety analysis  

SciTech Connect (OSTI)

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

80

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

SciTech Connect (OSTI)

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

WILLIAMS, J.C.

2003-11-15T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

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

Integrated Safety Management at the Activity Level: Work Planning and Control, Final Report Analysis of Integrated Safety Management at the Activity Level: Work Planning and...

82

Rapid Exploitation and Analysis of Documents  

SciTech Connect (OSTI)

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

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

2011-11-28T23:59:59.000Z

83

Nuclear Safety Delegations for Documented Safety Analysis Approval Criteria Review and Approach Document - January 8, 2015  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't32DepartmentWells517 Federali NOT

84

Safety Bulletin  

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

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...

85

ASD Facility Hazard Analysis Document - Building 382  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical

86

ASD Facility Hazard Analysis Document - Building 400  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical Equipment

87

ASD Facility Hazard Analysis Document - Building 411  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical1 Equipment

88

ASD Facility Hazard Analysis Document - Building 412  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical1 Equipment2

89

ASD Facility Hazard Analysis Document - Building 413  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical1

90

ASD Facility Hazard Analysis Document - Building 420  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical120 Equipment

91

ASD Facility Hazard Analysis Document - EAA  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical120

92

340 Waste handling Facility Hazard Categorization and Safety Analysis  

SciTech Connect (OSTI)

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

T. J. Rodovsky

2010-10-25T23:59:59.000Z

93

Document  

Office of Environmental Management (EM)

the safety, security, and reliability of existing nuclear warheads without the use of nuclear testing. Throughout the 1990s, DOE also took steps to consolidate the Complex from...

94

Document Reviews  

Broader source: Energy.gov [DOE]

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

95

Mechanistic facility safety and source term analysis  

SciTech Connect (OSTI)

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

PLYS, M.G.

1999-06-09T23:59:59.000Z

96

Rankine bottoming cycle safety analysis. Final report  

SciTech Connect (OSTI)

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

97

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

SciTech Connect (OSTI)

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

NONE

1986-12-01T23:59:59.000Z

98

Oak Ridge National Laboratory site data for safety-analysis report  

SciTech Connect (OSTI)

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

99

Office of Environment, Safety and Health Assessments Protocol...  

Office of Environmental Management (EM)

National Nuclear Security Administration CRAD, Radiological Controls - Idaho Accelerated Retrieval Project Phase II CRAD, Preliminary Documented Safety Analysis - July 25,...

100

Hazard Analysis Database report  

SciTech Connect (OSTI)

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

Niemi, B.J.

1997-08-12T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

Hazard analysis results report  

SciTech Connect (OSTI)

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

Niemi, B.J., Westinghouse Hanford

1996-09-30T23:59:59.000Z

102

Development of a pilot safety information document (PSID) for the replacement of radioactive liquid waste treatment facility at Los Alamos National Laboratory  

E-Print Network [OSTI]

Based on recent decisions made by Los Alamos National Laboratory concerning the development of site-wide National Environmental Policy Act documents, an effort was undertaken to develop a Pilot Safety Information Document (PSID) for the replacement...

Selvage, Ronald Derek

1995-01-01T23:59:59.000Z

103

242-A evaporator safety analysis report  

SciTech Connect (OSTI)

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

104

document  

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

to Document 07123.04.Final Page 28 the surface during pipe connections with conventional rotary rigs. A major contribution of fast drilling and less rig time for CT drilling is...

105

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

SciTech Connect (OSTI)

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

106

auditable safety analysis: Topics by E-print Network  

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

of environmental and safety analysis of fusion reactors MIT - DSpace Summary: This report summarizes the progress made between October 1976 and September 1977 in studies of...

107

Safety analysis of in-use vehicle wrapping cylinder  

Broader source: Energy.gov [DOE]

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

108

Monthly Analysis of Electrical Safety Occurrences – June 2011  

Broader source: Energy.gov [DOE]

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

109

Monthly Analysis of Electrical Safety Occurrences – August 2011  

Broader source: Energy.gov [DOE]

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

110

Monthly Analysis of Electrical Safety Occurrences – July 2011  

Broader source: Energy.gov [DOE]

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

111

Issues affecting advanced passive light-water reactor safety analysis  

SciTech Connect (OSTI)

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01T23:59:59.000Z

112

Issues affecting advanced passive light-water reactor safety analysis  

SciTech Connect (OSTI)

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-08-01T23:59:59.000Z

113

Hazard screening application guide. Safety Analysis Report Update Program  

SciTech Connect (OSTI)

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

114

Preparation of Nonreactor Nuclear Facility Documented Safety Analysis  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy1.pdfMarket |21,- EA-1999-07Disposition Project VolumecoolingThe

115

Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis -  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - SeptemberMicroneedles for medical point07.06 Agenda 2006UptakeDOE Directives,

116

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTie Ltd:June 2015 <Ones |Laboratory, JuneDid y ou Qualityandof

117

CRAD, Preliminary Documented Safety Analysis - July 25, 2014 (IEA CRAD  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTie Ltd:June 2015 <Ones |Laboratory, JuneDid31-2, REV. 0) |

118

Document  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to UserProduct:Directives Templates The OfficeDitchDivya401 Federal

119

Document  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to UserProduct:Directives Templates The OfficeDitchDivya401 Federal3623

120

PAT-1 safety analysis report addendum.  

SciTech Connect (OSTI)

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

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

2010-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

Stage Right operational safety analysis and evaluation of Pantex personnel operations  

SciTech Connect (OSTI)

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

Rountree, S.L.K.; Whitehurst, H.O.; Tomlin, E.H.; Restrepo, L.F. [Sandia National Labs., Albuquerque, NM (United States); White, J. [Sandia National Labs., Albuquerque, NM (United States)]|[Intera, Albuquerque, NM (United States)

1995-01-01T23:59:59.000Z

122

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

SciTech Connect (OSTI)

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

123

Safety and Environment Considerations and Analysis Safety and environmental issues are being considered up front in the APEX project as  

E-Print Network [OSTI]

Safety and Environment Considerations and Analysis Safety and environmental issues are being considered up front in the APEX project as designs evolve so that the goal of safety and environmental attractiveness is realized. Designing safety into the concepts as was done in the ITER project [1] results

California at Los Angeles, University of

124

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

SciTech Connect (OSTI)

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

125

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

SciTech Connect (OSTI)

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

Unwin, Stephen D.; Seiple, Timothy E.

2009-05-28T23:59:59.000Z

126

Document  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement Analysis,/U90 Federal

127

Document  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement Analysis,/U90 Federal2401

128

Document  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement Analysis,/U90

129

Document  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement Analysis,/U90federal

130

Document  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of theOFFICEACME |Supplement Analysis,/U90federal

131

Document:  

Office of Legacy Management (LM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545*. . : '* FEB1 Rev.DATA DECISION ANALYSIS:

132

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

SciTech Connect (OSTI)

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

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

2006-05-18T23:59:59.000Z

133

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

SciTech Connect (OSTI)

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

Milliken, N.J., Westinghouse Hanford

1996-08-01T23:59:59.000Z

134

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

SciTech Connect (OSTI)

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

Shedrow, C.B.

1999-11-29T23:59:59.000Z

135

Facility Safety  

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

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

136

C:\\Documents and Settings\\lm.SU-4155CABB2178\\Desktop\\LabSafety\\SA04002 Evaluation.DOC Date: November 9, 2004  

E-Print Network [OSTI]

C:\\Documents and Settings\\lm.SU-4155CABB2178\\Desktop\\LabSafety\\SA04002 Evaluation.DOC Date cultures capable of producing Alpha-Hemolysin Toxin be possessed by Dr. #12;C:\\Documents and Settings\\lm

Movileanu, Liviu

137

Safety analysis report for packaging (onsite) steel drum  

SciTech Connect (OSTI)

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

138

An Investigation of the Latent Semantic Analysis Technique for Document Retrieval  

E-Print Network [OSTI]

An Investigation of the Latent Semantic Analysis Technique for Document Retrieval STUDENT PROJECT;_________________________________________________________________________ An Investigation of the Latent Semantic Analysis Technique for Document Retrieval. Report by: David Mugo Page 2. These term-matching techniques have always relied on matching query terms with document terms to retrieve

139

A new DOE standard for transuranic waste nuclear safety analysis  

SciTech Connect (OSTI)

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

140

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

SciTech Connect (OSTI)

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

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

1992-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

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

1992-01-01T23:59:59.000Z

142

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

SciTech Connect (OSTI)

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

NONE

1994-10-01T23:59:59.000Z

143

RISMC ADVANCED SAFETY ANALYSIS WORKING PLAN – FY 2015 – FY 2019  

SciTech Connect (OSTI)

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 stakeholder’s use of the ASAP research and development (R&D) products. 2. An identification of likely end users and pathway to adoption of enhanced tools by the end-users. 3. A proposed set of practical and achievable “use case” demonstrations. 4. A proposed plan to address ASAP verification and validation (V&V) needs. 5. A proposed schedule for the multi-year ASAP.

Szilard, Ronaldo H; Smith, Curtis L

2014-09-01T23:59:59.000Z

144

Surveillance Analysis Computer System (SACS) software design document (SDD)  

SciTech Connect (OSTI)

This document contains the Software Design Description for Phase II of the SACS project, and Impact Level 3Q system

Glasscock, J.A.

1995-09-01T23:59:59.000Z

145

FAQS Gap Analysis Qualification Card – Criticality Safety  

Broader source: Energy.gov [DOE]

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

146

Safety Analysis Of Automated Highway Systems  

E-Print Network [OSTI]

Lee. Towards an automated fmea assis- tant. In Applicationsmodes and effects analysis (FMEA) is employed to determineof multiple failures. ) FMEA was developed Potential Part

Leveson, Nancy G.

1997-01-01T23:59:59.000Z

147

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

SciTech Connect (OSTI)

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

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

1999-01-15T23:59:59.000Z

148

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy1.pdfMarket |21,-CommitteeItems at

149

analysis documents volume: Topics by E-print Network  

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

document-level discourse parser. Our parser builds a discourse tree by applying an optimal parsing of discourse parsing effectively. A set of empirical evaluations over two...

150

Safety Analysis for Packaging Steel Banded Wooden Shipping Containers  

SciTech Connect (OSTI)

This safety analysis report for packaging describes the steel banded wooden shipping containers, which are certified as Type AF packagings. The authorized payload for these containers is unirradiated, slightly enriched, uranium ingots, billets, extrusions, and scrap materials. The amount of uranium in the containers will not exceed the LSA-II material requirements as defined in 49 CFR 173.403.

FERRELL, P.C.

2000-12-05T23:59:59.000Z

151

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

SciTech Connect (OSTI)

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

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

1986-02-01T23:59:59.000Z

152

Nuclear criticality safety guide  

SciTech Connect (OSTI)

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

Pruvost, N.L.; Paxton, H.C. [eds.] [eds.

1996-09-01T23:59:59.000Z

153

Nuclear criticality safety tools in the Chernobyl-4 accident analysis  

SciTech Connect (OSTI)

The collaboration with the Italian Safety Authority (DISP), started in July 1986, has the aim of studying, from a neutronic point of view, the possible initiator event and the accident dynamics in unit four of the Chernobly nuclear power plant. This report was produced within the framework of that collaboration. A main condition of the present work was making use of standard calculational methods employed in nuclear criticality safety analysis. This means that the neutron multiplication factor calculation should be made with the modules and the cross-section libraries of the SCALE system or in any case with some KENO IV version and the burnup calculation with the ORIGEN code.

Landeyro, P.A.

1988-01-01T23:59:59.000Z

154

DOE Standard Integration Of Environment,Safety, and Health Into...  

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

of Energy (DOE) approved methodology for preparing a Documented Safety Analysis (DSA) for decommissioning of nuclear facilities, as well as environmental restoration activities...

155

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

SciTech Connect (OSTI)

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

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

1992-02-01T23:59:59.000Z

156

SCALE 6: Comprehensive Nuclear Safety Analysis Code System  

SciTech Connect (OSTI)

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

157

Computational Methods for Sensitivity and Uncertainty Analysis in Criticality Safety  

SciTech Connect (OSTI)

Interest in the sensitivity methods that were developed and widely used in the 1970s (the FORSS methodology at ORNL among others) has increased recently as a result of potential use in the area of criticality safety data validation procedures to define computational bias, uncertainties and area(s) of applicability. Functional forms of the resulting sensitivity coefficients can be used as formal parameters in the determination of applicability of benchmark experiments to their corresponding industrial application areas. In order for these techniques to be generally useful to the criticality safety practitioner, the procedures governing their use had to be updated and simplified. This paper will describe the resulting sensitivity analysis tools that have been generated for potential use by the criticality safety community.

Broadhead, B.L.; Childs, R.L.; Rearden, B.T.

1999-09-20T23:59:59.000Z

158

Safety assessment document for the environmental test complex (Building 834) at Site 300  

SciTech Connect (OSTI)

A safety assessment was performed to determine if accidents occurring at the 834 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (HE) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire. Seismic and extreme wind (including missiles) analyses indicate that the buildings are basically sound. (However, there are a few recommendations to further enhance the structural integrity of these facilities). Additional lightning protection for these facilities is being installed. These buildings are located high above the dry creek bed so that a flood is improbable. A criticality or a high explosive detonation involving plutonium is very remote since the radioactive materials are encased and plutonium and HE are not permitted concurrently in the same area at Site 300. (The exceptions to this policy are that explosive actuating devices are sometimes located in assemblies containing fissile materials. However, a planned or accidental actuation will not effect the safe containment of the fissile material within the assembly). Even though the possibility of an HE explosion involving uranium and beryllium is remote, the off-site lung doses were calculated and found to be below the accepted standards. It was determined that a fire was unlikely due to the low fire loading and the absence of ignition sources. It was also determined that the consequences of any accidents were reduced by the remote location of these facilities, their design, and by administrative controls.

Odell, B.N.; Pfeifer, H.E.

1981-03-03T23:59:59.000Z

159

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

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

and systems analysis features are applicable to other liquid-metal cooled reactor concepts. Applications Safety analysis of fast reactors Simulations for...

160

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

SciTech Connect (OSTI)

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

Note: This page contains sample records for the topic "documented 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

Safety analysis report vitrified high level waste type B shipping cask  

SciTech Connect (OSTI)

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

NONE

1995-03-01T23:59:59.000Z

162

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

SciTech Connect (OSTI)

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

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

1991-07-01T23:59:59.000Z

163

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

164

Underground Test Area Subproject Phase I Data Analysis Task. Volume VIII - Risk Assessment Documentation Package  

SciTech Connect (OSTI)

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

None

1996-12-01T23:59:59.000Z

165

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

SciTech Connect (OSTI)

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

None

1996-12-01T23:59:59.000Z

166

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.

167

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.

168

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.

169

Safety and Environmental Considerations and Analysis APEX Interim Report November, 1999  

E-Print Network [OSTI]

Safety and Environmental Considerations and Analysis APEX Interim Report November, 1999 14-1 CHAPTER 14: SAFETY AND ENVIRONMENT CONSIDERATIONS AND ANALYSIS Contributors Lead Author: K.A. McCarthy H. Khater L. C. Cadwallader B. J. Merrill R.L. Moore D.A. Petti S. T. Schuetz #12;Safety and Environmental

California at Los Angeles, University of

170

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

E-Print Network [OSTI]

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon tries to assure the systems' safety through performing various safety analysis techniques ­ FTA (Fault was KNICS(Korea Nuclear Instrumentation and Control System) RPS(Reactor Protection System). · Prototype

171

ASD Facility Hazard Analysis Document - 400-EAA Mechanical Mezzanine  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical Mezzanine

172

ASD Facility Hazard Analysis Document - Building 400-EAA  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical

173

Retrieval of Patent Documents from Heterogeneous Sources using Ontologies and Similarity Analysis  

E-Print Network [OSTI]

Retrieval of Patent Documents from Heterogeneous Sources using Ontologies and Similarity Analysis related to the patent system. Patents and related documents are siloed into multiple heterogeneous sources in the bio-patent domain with a collection of 1150 patents and 30 court cases. Keywords-Ontology, Patent

Stanford University

174

Technical safety requirements (TSR) for waste receiving and processing (WRAP) facility  

SciTech Connect (OSTI)

The scope of this TSR document is based on the WRAP Final Safety Analysis Report (HNF-SD-W026-SAR-002) and supporting documents. The administrative controls set forth in this TSR document are derived from the WRAP Final Safety Analysis Report.

Weidert, J.R.

1997-11-18T23:59:59.000Z

175

DOE's Approach to Nuclear Facility Safety Analysis and Management  

Broader source: Energy.gov [DOE]

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

176

Fast Flux Test Facility final safety analysis report. Amendment 73  

SciTech Connect (OSTI)

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

177

The Zion integrated safety analysis for NUREG-1150  

SciTech Connect (OSTI)

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

178

Complete Experiment Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed New SubstationCleanCommunity Involvement andMISR, and4 N Overview of HowComplete

179

Complete Experiment Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed New SubstationCleanCommunity Involvement andMISR, and4 N Overview of

180

WIPP Documents - Nuclear Safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert SouthwestTechnologies | Blandine Jerome VelenciaNewsNewsFederal RegulationsLabel

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


181

Acceptable NSLS Safety Documentation  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)ProductssondeadjustsondeadjustAbout the Building Technologies Office Workshop WorkingAccelerators

182

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

SciTech Connect (OSTI)

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

NONE

1994-10-01T23:59:59.000Z

183

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

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

184

Analysis of Integrated Safety Management at the Activity Level: Work  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergy Cooperation |South42.2 (April 2012) 1 DocumentationAnalysis of Crossover Points forNEPA

185

Exploratory Analysis of Concept and Document Spaces with Connectionist Networks  

E-Print Network [OSTI]

. Exploratory analysis is an area of increasing interest in the computational linguistics arena. Pragmatically speaking, exploratory analysis may be paraphrased as natural language processing by means of analyzing large corpora of text. Concerning the analysis, appropriate means are statistics, on the one hand, and artificial neural networks, on the other hand. As a challenging application area for exploratory analysis of text corpora we may certainly identify text databases, be it information retrieval or information filtering systems. With this paper we present recent findings of exploratory analysis based on both statistical and neural models applied to legal text corpora. Concerning the artificial neural networks, we rely on a model adhering to the unsupervised learning paradigm. This choice appears naturally when taking into account the specific properties of large text corpora where one is faced with the fact that input-output-mappings as required by supervised learning models ca...

1999-01-01T23:59:59.000Z

186

New enhancements to SCALE for criticality safety analysis  

SciTech Connect (OSTI)

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

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

1995-09-01T23:59:59.000Z

187

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

E-Print Network [OSTI]

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon through safety analy- sis is strongly mandated for safety-critical systems. Nuclear plant protection, NuFTA, for nuclear plant protection systems. NuFTA mechanically constructs a software fault tree

188

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

189

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

190

Safety analysis report for packaging (Onsite) for the Hanford Ecorok packaging  

SciTech Connect (OSTI)

This safety analysis report for packaging approves the Hanford Ecorok packaging for shipping contaminated water purification filters from K Basins to the Central Waste Complex.

Mercado, M.S.

1996-02-23T23:59:59.000Z

191

Safety analysis report for packaging (onsite) doorstop samplecarrier system  

SciTech Connect (OSTI)

The Doorstop Sample Carrier System consists of a Type B certified N-55 overpack, U.S. Department of Transportation (DOT) specification or performance-oriented 208-L (55-gal) drum (DOT 208-L drum), and Doorstop containers. The purpose of the Doorstop Sample Carrier System is to transport samples onsite for characterization. This safety analysis report for packaging (SARP) provides the analyses and evaluation necessary to demonstrate that the Doorstop Sample Carrier System meets the requirements and acceptance criteria for both Hanford Site normal transport conditions and accident condition events for a Type B package. This SARP also establishes operational, acceptance, maintenance, and quality assurance (QA) guidelines to ensure that the method of transport for the Doorstop Sample Carrier System is performed safely in accordance with WHC-CM-2-14, Hazardous Material Packaging and Shipping.

Obrien, J.H.

1997-02-24T23:59:59.000Z

192

ARM - Publications: Science Team Meeting Documents: Analysis of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD) by Microtops AtmosphericApplication andAn AssessmentARMArctic Facility forAnalysis ofground-based

193

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

SciTech Connect (OSTI)

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

194

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

SciTech Connect (OSTI)

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

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

2008-11-01T23:59:59.000Z

195

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

SciTech Connect (OSTI)

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

Garvin, L.J.

1995-11-01T23:59:59.000Z

196

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

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

197

Battery Lifetime Analysis and Simulation Tool (BLAST) Documentation  

SciTech Connect (OSTI)

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

Neubauer, J.

2014-12-01T23:59:59.000Z

198

Peer review of the National Transportation Safety Board structural analysis of the I-35W bridge collapse.  

SciTech Connect (OSTI)

The Engineering Sciences Center at Sandia National Laboratories provided an independent peer review of the structural analysis supporting the National Transportation Safety Board investigation of the August 1, 2007 collapse of the I-35W Bridge in Minneapolis. The purpose of the review was to provide an impartial critique of the analysis approach, assumptions, solution techniques, and conclusions. Subsequent to reviewing numerous supporting documents, a SNL team of staff and management visited NTSB to participate in analysis briefings, discussions with investigators, and examination of critical elements of the bridge wreckage. This report summarizes the opinion of the review team that the NTSB analysis effort was appropriate and provides compelling supporting evidence for the NTSB probable cause conclusion.

Gwinn, Kenneth West; Redmond, James Michael; Wellman, Gerald William

2008-10-01T23:59:59.000Z

199

Interface design of VSOP'94 computer code for safety analysis  

SciTech Connect (OSTI)

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Natsir, Khairina, E-mail: yenny@batan.go.id; Andiwijayakusuma, D.; Wahanani, Nursinta Adi [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Yazid, Putranto Ilham [Center for Nuclear Technology, Material and Radiometry- National Nuclear Energy Agency, Jl. Tamansari No.71, Bandung 40132 (Indonesia)

2014-09-30T23:59:59.000Z

200

Criticality safety analysis on fissile materials in Fukushima reactor cores  

SciTech Connect (OSTI)

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

Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA 94720 (United States); Hirano, Fumio [Japan Atomic Energy Agency, Geological Isolation Research and Development Directorate, Tokai-mura, Ibaraki 319-1194 (Japan)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

NEW - DOE O 420.1 Chg 1, Facility Safety  

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

The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. This Page Change is limited in scope to changes necessary to invoke DOE-STD-1104, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Document, and revised DOE-STD-3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis as required methods. DOE O 420.1C Chg 1, dated 2-27-15, cancels DOE O 420.1C, dated 12-4-12.

202

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation  

SciTech Connect (OSTI)

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

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

2009-11-01T23:59:59.000Z

203

Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings  

SciTech Connect (OSTI)

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

204

Setting clear expectations for safety basis development  

SciTech Connect (OSTI)

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

205

TASSEL 3.0 Universal Network Enabled Analysis Kit (UNEAK) pipeline documentation  

E-Print Network [OSTI]

1 TASSEL 3.0 Universal Network Enabled Analysis Kit (UNEAK) pipeline documentation Authors: Fei Lu.............................................................................................................................. 8 Introduction The UNEAK is the non-reference Genotyping by Sequencing (GBS) SNP calling pipeline the command line in the following format (Linux or Mac operating system; for Windows use run_pipeline

Buckler, Edward S.

206

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

SciTech Connect (OSTI)

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

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2014-02-12T23:59:59.000Z

207

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

208

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

SciTech Connect (OSTI)

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

Not Available

1993-01-01T23:59:59.000Z

209

New Methods and Tools to Perform Safety Analysis within RISMC  

SciTech Connect (OSTI)

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

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

2013-11-01T23:59:59.000Z

210

340 waste handling facility interim safety basis  

SciTech Connect (OSTI)

This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

VAIL, T.S.

1999-04-01T23:59:59.000Z

211

340 Waste handling facility interim safety basis  

SciTech Connect (OSTI)

This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

Stordeur, R.T.

1996-10-04T23:59:59.000Z

212

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

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

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

2011-12-19T23:59:59.000Z

213

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

SciTech Connect (OSTI)

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

214

Safety Analysis Report: X17B2 beamline Synchrotron Medical Research Facility  

SciTech Connect (OSTI)

This report contains a safety analysis for the X17B2 beamline synchrotron medical research facility. Health hazards, risk assessment and building systems are discussed. Reference is made to transvenous coronary angiography. (LSP)

Gmuer, N.F.; Thomlinson, W.

1990-02-01T23:59:59.000Z

215

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

216

Use of Fault Tree Analysis for Automotive Reliability and Safety Analysis  

SciTech Connect (OSTI)

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

217

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

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the U.S. Department of Energy (DOE), on the Safety Analysis Report for Packaging, Model 9975, Revision 0, dated January 2008 (S-SARP-G-00003, the SARP). The review includes an evaluation of the SARP, with respect to the requirements specified in 10 CFR 71, and in International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9975-96 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The safety analysis of the Model 9975-85 Packaging is documented in the Safety Analysis Report for Packaging, Model 9975, B(M)F-85, Revision 0, dated December 2003. The Model 9975-85 Package is certified by DOE Certificate of Compliance (CoC) package identification number, USA/9975/B(M)F-85, for the transportation of Type B quantities of uranium metal/oxide, {sup 238}Pu heat sources, plutonium/uranium metals, plutonium/uranium oxides, plutonium composites, plutonium/tantalum composites, {sup 238}Pu oxide/beryllium metal.

West, M

2009-05-22T23:59:59.000Z

218

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

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

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

219

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

220

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

SciTech Connect (OSTI)

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

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

2011-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

Smith, Curtis L; Mandelli, Diego; Zhegang Ma

2014-11-01T23:59:59.000Z

222

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

223

Implementation Guide for Use in Developing Technical Safety Requirements  

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

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

2010-11-03T23:59:59.000Z

224

Implementation Guide for Use in Developing Technical Safety Requirements  

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

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

2010-11-03T23:59:59.000Z

225

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy1.pdfMarket |21,- EA-1999-07Disposition Project Volume 2PREMIUM

226

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy1.pdfMarket |21,- EA-1999-07Disposition Project Volume 2PREMIUM

227

Factors Analysis on Safety of Indoor Air Quality  

E-Print Network [OSTI]

. Handbook on Review and Detection of Indoor Environment [M]. Beijing: Mechanical Industry Press, 2003: 1-5.(In Chinese) [2] Pan Xiaochuan. Review on Indoor Air Pollution and Its Harmfulness to Health [J]. Chin. Prev. Med., 2002,3(3):167-169 (in... of Urban Construction, Nanhua University, Hengyang, P.R.China hunanluoqinghai@163.com Abstract: Influence factors on safety of indoor air quality (IAQ) were analyzed in this paper. Some regeneration compositions resulting from potential indoor...

Luo, Q.; Liu, Z.; Xiong, J.

2006-01-01T23:59:59.000Z

228

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

SciTech Connect (OSTI)

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

229

TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS  

SciTech Connect (OSTI)

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

KRIPPS, L.J.

2006-07-31T23:59:59.000Z

230

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

231

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

232

SACRD: a data base for fast reactor safety computer codes, operational procedures  

SciTech Connect (OSTI)

SACRD (Safety Analysis Computerized Reactor Data) is a data base of nondesign-related information used in computer codes for fast reactor safety analyses. This document reports the procedures used in SACRD to help assure a reasonable level of integrity of the material contained in the data base. It also serves to document much of the computer software used with the data base.

Forsberg, V.M.; Arwood, J.W.; Greene, N.M.; Raiford, G.B.

1980-09-01T23:59:59.000Z

233

Safety System Oversight Staffing Analysis (Instructions, Blank Sheet and  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalancedDepartmentRestrictionsExample Sheet) | Department of Energy Safety

234

System Design and the Safety Basis  

SciTech Connect (OSTI)

The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination & decommissioning (D&D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities.

Ellingson, Darrel

2008-05-06T23:59:59.000Z

235

Invited Contribution to Q 76: The Use of Risk Analysis to Support Dam Safety Decisions and Management  

E-Print Network [OSTI]

Decisions and Management DRAFT FOR REVIEW ONLY Portfolio Risk Assessment: A Tool for Managing Dam SafetyICOLD 20th Congress Invited Contribution to Q 76: The Use of Risk Analysis to Support Dam Safety in the Context of the Owner's Business David S. Bowles Professor and Director, Institute for Dam Safety Risk

Bowles, David S.

236

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

SciTech Connect (OSTI)

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

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

1992-07-01T23:59:59.000Z

237

Review and Analysis of Development of "Safety by Design" Requirements  

SciTech Connect (OSTI)

This report, the deliverable for Task 4 of the NA-243 Safeguards by Design Work Plan for Fiscal Year 2009, develops the lessons to be learned for the institutionalization of Safeguards By Design (SBD) from the Department of Energy (DOE) experience developing and implementing DOE-STD-1189, Integration of Safety into the Design Process. This experience was selected for study because of the similarity of the challenges of integrating safety and safeguards into the design process. Development of DOE-STD-1189 began in January 2006 and the standard was issued for implementation in March 2008. The process was much more time consuming than originally anticipated and might not have come to fruition had senior DOE management been less committed to its success. Potentially valuable lessons can be learned from both the content and presentation of the integration approach in DOE-STD-1189 and from the DOE experience in developing and implementing DOE-STD-1189. These lessons are important because the instutionalization of SBD does not yet appear to have the level of senior management commitment afforded development and implementation of DOE-STD-1189.

Vance, Scott A.; Hockert, John

2009-10-20T23:59:59.000Z

238

Analysis of Fundamental NIST Sphere Experiments Related to Criticality Safety  

SciTech Connect (OSTI)

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

239

OCCUPATIONAL & ENVIRONMENTAL SAFETY SERVICES C:\\Documents and Settings\\sysdev\\Local Settings\\Temporary Internet Files\\OLK412\\Lab Equipment Release -New Format.doc  

E-Print Network [OSTI]

, autoclaves, centrifuges, refrigerators, freezers, incubators, BioSafety cabinets, and analytical equipment, refrigerators, freezers, incubators, etc. #12;OCCUPATIONAL & ENVIRONMENTAL SAFETY SERVICES C

Krovi, Venkat

240

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

SciTech Connect (OSTI)

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

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

1996-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

242

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

SciTech Connect (OSTI)

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

NONE

1994-10-01T23:59:59.000Z

243

FAQS Gap Analysis Qualification Card – Nuclear Explosive Safety Study  

Broader source: Energy.gov [DOE]

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

244

Tank farms criticality safety manual  

SciTech Connect (OSTI)

This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type.

FORT, L.A.

2003-03-27T23:59:59.000Z

245

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

Broader source: Energy.gov [DOE]

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

246

Fusion Engineering and Design 38 (1997) 189218 ARIES-RS safety design and analysis  

E-Print Network [OSTI]

Fusion Engineering and Design 38 (1997) 189­218 ARIES-RS safety design and analysis D. Steiner *, L Polytechnic Institute, Department of En6ironmental and Energy Engineering, JEC 5049, Troy NY 12180-3590, USA assessment indicates that the dose at the site boundary will be less than 1 rem per year. Thus, no sheltering

247

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

SciTech Connect (OSTI)

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

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

2005-08-08T23:59:59.000Z

248

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

249

FACILITY SAFETY (FS)  

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

FACILITY SAFETY (FS) OBJECTIVE FS.1 - (Core Requirement 7) Facility safety documentation in support of SN process operations,is in place and has been implemented that describes the...

250

Coiled Tubing Safety Manual  

SciTech Connect (OSTI)

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

Crow, W.

1999-04-06T23:59:59.000Z

251

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

SciTech Connect (OSTI)

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

252

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

SciTech Connect (OSTI)

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

253

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

SciTech Connect (OSTI)

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

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

1983-08-01T23:59:59.000Z

254

Criticality safety analysis of a borated-concrete absorber  

SciTech Connect (OSTI)

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

255

Supporting Document Strategic Plan  

E-Print Network [OSTI]

1 Supporting Document Strategic Plan 2013­2020 #12;2 Supporting Document Strategic Plan 2013 more critical to the University's future than was the case in 2005. The purpose of this document are summarised via a SWOT analysis in Appendix 1. This document should therefore be read in conjunction

Auckland, University of

256

FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS  

SciTech Connect (OSTI)

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

GARVIN, L J; JENSEN, M A

2004-04-13T23:59:59.000Z

257

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

SciTech Connect (OSTI)

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

258

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

SciTech Connect (OSTI)

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

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

259

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

SciTech Connect (OSTI)

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

260

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

SciTech Connect (OSTI)

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

Note: This page contains sample records for the topic "documented 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

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

E-Print Network [OSTI]

to coincide with the wave crest. Ferranti Offshore Systems, a crane manufacturer, markets a retrofittable system that has a hook-mounted control unit. This system has a rope hanging down for the personnel on the deck to control the heave compensation...DEVELOPMENT OF A SAFETY ANALYSIS SYSTEM FOR THE OFFSHORE PERSONNEL AND EQUIPMENT TRANSFER PROCESS A Thesis by MICHAEL GEORGE McKENNA Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment...

McKenna, Michael George

1988-01-01T23:59:59.000Z

262

Topaz II preliminary safety assessment  

SciTech Connect (OSTI)

The Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary safety assessment included a top level event tree, neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, and analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment, it appears that it will be possible to safely launch the Topaz II system in the U.S. with some possible system modifications. The principal system modifications will probably include design changes to preclude water flooded criticality and to assure intact reentry.

Marshall, A.C. (Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States)); Standley, V. (Air Force Phillips Laboratory, Albuquerque, New Mexico 87110 (United States)); Voss, S.S. (Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)); Haskin, E. (Department of Chemical and Nuclear Engineering Department, Institute for Nuclear Power Studies, University of New Mexico, Albuquerque, New Mexico 87110 (United States))

1993-01-10T23:59:59.000Z

263

Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks  

SciTech Connect (OSTI)

The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff 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 CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches.

Not Available

1988-01-01T23:59:59.000Z

264

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

SciTech Connect (OSTI)

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

265

RESEARCH SAFETY RADIATION SAFETY  

E-Print Network [OSTI]

RESEARCH SAFETY RADIATION SAFETY ENVIRONMENTAL PROGRAMS HAZARDOUS MATERIALS CONTROLLED SUBSTANCES INTEGRATED WASTE MANAGEMENT LABORATORY SAFETY AUDITS & COMPLIANCE BIOSAFETY and ENVIRONMENTAL HEALTH EMERGENCY MANAGEMENT and MISSION CONTINUITY FIRE PREVENTION and LIFE SAFETY GENERAL SAFETY TRAINING

266

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

SciTech Connect (OSTI)

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

267

Text Retrieval from Document Images based on Word Shape Analysis Chew Lim Tan, Weihua Huang, Sam Yuan Sung, Zhaohui Yu and Yi Xu  

E-Print Network [OSTI]

Text Retrieval from Document Images based on Word Shape Analysis Chew Lim Tan, Weihua Huang, Sam, Singapore 117543 Abstract In this paper, we propose a method of text retrieval from document images using a similarity measure based on word shape analysis. We directly extract image features instead of using optical

Tan, Chew Lim

268

Canister storage building hazard analysis report  

SciTech Connect (OSTI)

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

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

1997-07-01T23:59:59.000Z

269

Cold Vacuum Drying Facility hazard analysis report  

SciTech Connect (OSTI)

This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) hazard analysis to support the CVDF phase 2 safety analysis report (SAR), and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, and implements the requirements of US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports.

Krahn, D.E.

1998-02-23T23:59:59.000Z

270

Environmental Health and Safety  

E-Print Network [OSTI]

Environmental Health and Safety Approved by Document No. Version Date Replaces Page EHS EHS-FORM-072 1.0 15-May-2008 N/A 1 of 4 Laboratory Safety Orientation Checklist Name (Print) Department Supervisor Date (DD/MM/YY) A Laboratory Safety Orientation Checklist should be completed within one month

Shoubridge, Eric

271

SYSTEM SAFETY PROGRESS REPORT,  

E-Print Network [OSTI]

SYSTEM SAFETY PROGRESS REPORT, ALSEP Array E NO. ATM 1034 1 PAGE REV. NO. OF 3 DATE 26 July 1971 This A TM documents the progress of the System Safety Program for ALSEP Array E. -~/ Prepared by: · /~t:A~.., Approved by: W. · Lavin, Jr System Safety Engineer / /' J. P. ~/ es, Supervisor · , ALSEF Support

Rathbun, Julie A.

272

RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT  

SciTech Connect (OSTI)

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

KOZLOWSKI, S.D.

2007-05-30T23:59:59.000Z

273

Fusion integral experiments and analysis and the determination of design safety factors - I: Methodology  

SciTech Connect (OSTI)

The role of the neutronics experimentation and analysis in fusion neutronics research and development programs is discussed. A new methodology was developed to arrive at estimates to design safety factors based on the experimental and analytical results from design-oriented integral experiments. In this methodology, and for a particular nuclear response, R, a normalized density function (NDF) is constructed from the prediction uncertainties, and their associated standard deviations, as found in the various integral experiments where that response, R, is measured. Important statistical parameters are derived from the NDF, such as the global mean prediction uncertainty, and the possible spread around it. The method of deriving safety factors from many possible NDFs based on various calculational and measuring methods (among other variants) is also described. Associated with each safety factor is a confidence level, designers may choose to have, that the calculated response, R, will not exceed (or will not fall below) the actual measured value. An illustrative example is given on how to construct the NDFs. The methodology is applied in two areas, namely the line-integrated tritium production rate and bulk shielding integral experiments. Conditions under which these factors could be derived and the validity of the method are discussed. 72 refs., 17 figs., 4 tabs.

Youssef, M.Z.; Kumar, A.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States); Oyama, Y.; Maekawa, H. [Japan Atomic Energy Research Inst., Ibaraki (Japan)

1995-09-01T23:59:59.000Z

274

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

SciTech Connect (OSTI)

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

275

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

SciTech Connect (OSTI)

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

Not Available

1991-10-18T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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

277

Topaz II preliminary safety assessment  

SciTech Connect (OSTI)

The Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz 11 space nuclear power system. A preliminary safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safely assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary safety assessment included a top level event tree, neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, and analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment, it appears that it will be possible to safely launch the Topaz II system in the US with some possible system modifications. The principal system modifications will probably include design changes to preclude water flooded criticality and to assure intact reentry.

Marshall, A.C. (Sandia National Labs., Albuquerque, NM (United States)); Standley, V. (Air Force Phillips Laboratory, Albuquerque, NM (United States)); Voss, S.S. (Los Alamos National Lab., NM (United States)); Haskin, E. (New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering)

1992-01-01T23:59:59.000Z

278

Supervisors` orientation to occupational safety in DOE  

SciTech Connect (OSTI)

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

NONE

1993-10-01T23:59:59.000Z

279

UNCLASSIFIHED DEFENSE DOCUMENTATION CENTER  

E-Print Network [OSTI]

UNCLASSIFIHED AD 463473 DEFENSE DOCUMENTATION CENTER FOR SCIENTIFIC AND TECHNICAL INFORMATION .... John Barton Head OR Analysis Group R. H. Krolick Manager Applied Science Laboratory Prepared for the .J

Block, Marco

280

E-Print Network 3.0 - accident analysis documentation Sample...  

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

and Data Analysis, 2000). This has... Research Record 1487, 68-74. Grder, P., 1994. Bicycle accidents in Maine: an analysis. Transportation... at signalized intersections,...

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

282

SWANSEA UNIVERSITY OCCUPATIONAL HEALTH AND SAFETY POLICY  

E-Print Network [OSTI]

of the University, as laid down by the HSE "Successful Health and Safety Management" document (HSG65) is "to improve: Health and Safety Policy 2. Specific Policy Objectives 3. Organization of Health and Safety Management 4 Documentation Appendix A. Form EHSI B. Emergency Management Team 2 #12;1. Introduction: Health and Safety Policy

Martin, Ralph R.

283

ELECTRICAL POWER SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect (OSTI)

The purpose of this revision of the System Description Document (SDD) is to establish requirements that drive the design of the electrical power system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience are design engineers. This type of SDD leads and follows the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential to performing the design process. This SDD follows the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to this system are obtained from ''Project Functional and Operational Requirements'' (F&OR) (Siddoway, 2003). Other requirements to support the design process have been taken from higher level requirements documents such as ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), the fire hazards analyses, and the preclosure safety analysis. The above mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canori and Leitner 2003) requirements. This SDD includes several appendices with supporting information. Appendix B lists key system charts, diagrams, drawings, and lists; and Appendix C is a list of system procedures.

M. Maniyar

2004-06-22T23:59:59.000Z

284

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

SciTech Connect (OSTI)

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

Frappaolo, C. [Delphi Consulting Group, Boston, MA (United States)

1995-09-01T23:59:59.000Z

285

ASD Facility Hazard Analysis Document - 400-EAA Top of Storage Ring  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth (AOD)Productssondeadjustsondeadjust DocumentationARMStreamsUSBudget Advanced DISCLAIMERMechanical MezzanineTop

286

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

SciTech Connect (OSTI)

The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

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

2013-05-01T23:59:59.000Z

287

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

SciTech Connect (OSTI)

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

288

Safety analysis report for the use of hazardous production materials in photovoltaic applications at the National Renewable Energy Laboratory. Volume 2, Appendices  

SciTech Connect (OSTI)

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

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

1992-07-01T23:59:59.000Z

289

Plutonium Finishing Plant safety evaluation report  

SciTech Connect (OSTI)

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

290

Cold Vacuum Drying (CVD) Facility Hazards Analysis Report  

SciTech Connect (OSTI)

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

CROWE, R.D.

2000-08-07T23:59:59.000Z

291

Technical support document: Energy conservation standards for consumer products: Refrigerators and furnaces including: environmental impacts regulatory impact analysis  

SciTech Connect (OSTI)

The National Appliance Energy Conversation Act (NAECA) of 1987 (P.L. 100-12) establishes energy efficiency standards for 13 types of consumer products. The legislation requires the Department of Energy (DOE) to consider new or amended standards on these and other types of products at specified times. DOE is currently selecting standards for two types of products: refrigerators, refrigerator-freezers, and freezers; and small gas furnaces. This Technical Support Document presents the methodology, data and results from the analysis of the energy and economic impacts of the proposed standards. 8 refs., 39 figs., 135 tabs.

Not Available

1989-11-01T23:59:59.000Z

292

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

SciTech Connect (OSTI)

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

Dr. Bradley J Schrader

2010-10-01T23:59:59.000Z

293

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

SciTech Connect (OSTI)

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

Dr. Bradley J Schrader

2009-03-01T23:59:59.000Z

294

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

SciTech Connect (OSTI)

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

295

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

SciTech Connect (OSTI)

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

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

2004-12-01T23:59:59.000Z

296

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

297

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

SciTech Connect (OSTI)

The Department of Energy has assigned to Sandia National Laboratories the responsibility of producing a Safety Analysis Report (SAR) for the plutonium-dioxide fueled Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) proposed to be used in the Mars Science Laboratory (MSL) mission. The National Aeronautic and Space Administration (NASA) is anticipating a launch in fall of 2009, and the SAR will play a critical role in the launch approval process. As in past safety evaluations of MMRTG missions, a wide range of potential accident conditions differing widely in probability and seventy must be considered, and the resulting risk to the public will be presented in the form of probability distribution functions of health effects in terms of latent cancer fatalities. The basic descriptions of accident cases will be provided by NASA in the MSL SAR Databook for the mission, and on the basis of these descriptions, Sandia will apply a variety of sophisticated computational simulation tools to evaluate the potential release of plutonium dioxide, its transport to human populations, and the consequent health effects. The first step in carrying out this project is to evaluate the existing computational analysis tools (computer codes) for suitability to the analysis and, when appropriate, to identify areas where modifications or improvements are warranted. The overall calculation of health risks can be divided into three levels of analysis. Level A involves detailed simulations of the interactions of the MMRTG or its components with the broad range of insults (e.g., shrapnel, blast waves, fires) posed by the various accident environments. There are a number of candidate codes for this level; they are typically high resolution computational simulation tools that capture details of each type of interaction and that can predict damage and plutonium dioxide release for a range of choices of controlling parameters. Level B utilizes these detailed results to study many thousands of possible event sequences and to build up a statistical representation of the releases for each accident case. A code to carry out this process will have to be developed or adapted from previous MMRTG missions. Finally, Level C translates the release (or ''source term'') information from Level B into public risk by applying models for atmospheric transport and the health consequences of exposure to the released plutonium dioxide. A number of candidate codes for this level of analysis are available. This report surveys the range of available codes and tools for each of these levels and makes recommendations for which choices are best for the MSL mission. It also identities areas where improvements to the codes are needed. In some cases a second tier of codes may be identified to provide supporting or clarifying insight about particular issues. The main focus of the methodology assessment is to identify a suite of computational tools that can produce a high quality SAR that can be successfully reviewed by external bodies (such as the Interagency Nuclear Safety Review Panel) on the schedule established by NASA and DOE.

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

2006-09-01T23:59:59.000Z

298

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

SciTech Connect (OSTI)

A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations.

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

2013-10-01T23:59:59.000Z

299

Safety Basis Report  

SciTech Connect (OSTI)

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

R.J. Garrett

2002-01-14T23:59:59.000Z

300

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

SciTech Connect (OSTI)

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

Note: This page contains sample records for the topic "documented 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.


301

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

SciTech Connect (OSTI)

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

ULLAH, M K

2001-02-26T23:59:59.000Z

302

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

Energy Savers [EERE]

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

303

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

SciTech Connect (OSTI)

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

Not Available

1993-04-01T23:59:59.000Z

304

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

SciTech Connect (OSTI)

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

NONE

1999-03-01T23:59:59.000Z

305

Facility Safety  

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

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

1996-10-24T23:59:59.000Z

306

Facility Safety  

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

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

1995-11-16T23:59:59.000Z

307

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

SciTech Connect (OSTI)

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

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

2012-12-04T23:59:59.000Z

308

Technical documentation in support of the project-specific analysis for construction and operation of the National Ignition Facility  

SciTech Connect (OSTI)

This document provides information that supports or supplements the data and impact analyses presented in the National Ignition Facility (NIF) Project-Specific Analysis (PSA). The purposes of NIF are to achieve fusion ignition in the laboratory for the first time with inertial confinement fusion (ICF) technology and to conduct high- energy-density experiments ins support of national security and civilian application. NIF is an important element in the DOE`s science-based SSM Program, a key mission of which is to ensure the reliability of the nation`s enduring stockpile of nuclear weapons. NIF would also advance the knowledge of basic and applied high-energy- density science and bring the nation a large step closer to developing fusion energy for civilian use. The NIF PSA includes evaluations of the potential environmental impacts of constructing and operating the facility at one of five candidate site and for two design options.

Lazaro, M.A.; Vinikour, W. [Argonne National Lab., IL (United States). Environmental Assessment Div.; Allison, T. [Argonne National Lab., IL (United States). Decision and Information Sciences Div.] [and others

1996-09-01T23:59:59.000Z

309

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

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

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

310

Spent Nuclear Fuel project integrated safety management plan  

SciTech Connect (OSTI)

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

Daschke, K.D.

1996-09-17T23:59:59.000Z

311

Nuclear Engineer (Criticality Safety)  

Broader source: Energy.gov [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...

312

ENVIRONMENT, SAFETY & HEALTH DIVISION Chapter 10: Laser Safety  

E-Print Network [OSTI]

ENVIRONMENT, SAFETY & HEALTH DIVISION Chapter 10: Laser Safety Laser Service Subcontractor Work is unavailable), and the subcontractor. 2 Procedures The LSO will review the work plans, provide safety oversight that on-site work will be done that requires Site-specific safety plan (SSSP) and job safety analysis

Wechsler, Risa H.

313

ISV safety, processing, and starter path issues  

SciTech Connect (OSTI)

Numerous experiments and studies related to safety concerns in the in situ vitrification (ISV) process have been conducted at Pacific Northwest Laboratory. Topics of interest include (1) combustible inclusions, (2) sealed containers, (3) radiant heat surge, (4) electrical shock, (5) general risk analysis, and (6) Pu criticality. The data and analyses are those used for the initial ISV development and subsequent improvement; the majority was performed in 1987 or earlier. The purpose of this report is to document these analyses for reference purposes; knowledge gained more recently is, or will be, incorporated in other documents. 33 refs., 1 fig., 9 tabs.

Hilliard, D K; Kindle, C H [comps.

1991-04-01T23:59:59.000Z

314

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

315

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

SciTech Connect (OSTI)

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

316

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

SciTech Connect (OSTI)

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

317

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

SciTech Connect (OSTI)

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

318

Technical Report Documentation Page 1. Report No.  

E-Print Network [OSTI]

Technical Report Documentation Page 1. Report No. FHWA/TX-04/0-4266-2 2. Government Accession No. 3 significant changes in traffic operation and safety. A review of the design standard documents shows Statement No restrictions. This document is available to the public through the National Technical

Texas at Austin, University of

319

Technical Report Documentation Page 1. Report No.  

E-Print Network [OSTI]

Technical Report Documentation Page 1. Report No. FHWA/TX-04/5-1873-01-1 2. Government Accession No to improve safety and facilitate traffic flow. Finally, they documented these design considerations. This document is available to the public through the National Technical Information Service, Springfield

Texas at Austin, University of

320

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

SciTech Connect (OSTI)

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

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

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

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

SciTech Connect (OSTI)

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

322

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

SciTech Connect (OSTI)

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

Bloyd, C.; Camp, J.; Conzelmann, G. [and others

1996-12-01T23:59:59.000Z

323

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

SciTech Connect (OSTI)

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

324

Review and Approval of Nuclear Facility Safety Basis Documents (Documented Safety Analyses and Technical Safety Requirements)  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromComments onReplyofRetiring ProcurementReturn onRevenueNOT

325

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

SciTech Connect (OSTI)

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

326

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

SciTech Connect (OSTI)

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

United States. Bonneville Power Administration.

2005-11-01T23:59:59.000Z

327

Literature review of environmental qualification of safety-related electric cables: Literature analysis and appendices. Volume 2  

SciTech Connect (OSTI)

In support of the US NRC Environmental Qualification (EQ) Research Program, a literature review was performed to identify past relevant work that could be used to help fully or partially resolve issues of interest related to the qualification of low-voltage electric cable. A summary of the literature reviewed is documented in Volume 1 of this report. In this, Volume 2 of the report, dossiers are presented which document the issues selected for investigation in this program, along with recommendations for future work to resolve the issues, when necessary. The dossiers are based on an analysis of the literature reviewed, as well as expert opinions. This analysis includes a critical review of the information available from past and ongoing work in thirteen specific areas related to EQ. The analysis for each area focuses on one or more questions which must be answered to consider a particular issue resolved. Results of the analysis are presented, along with recommendations for future work. The analysis is documented in the form of a dossier for each of the areas analyzed.

Lofaro, R.; Bowerman, B.; Carbonaro, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

1996-04-01T23:59:59.000Z

328

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

329

Canister storage building hazard analysis report  

SciTech Connect (OSTI)

This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis was performed in accordance with the DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', and meets the intent of HNF-PRO-704, ''Hazard and Accident Analysis Process''. This hazard analysis implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports''.

POWERS, T.B.

1999-05-11T23:59:59.000Z

330

Nuclear criticality safety bounding analysis for the in-tank-precipitation (ITP) process, impacted by fissile isotopic weight fractions  

SciTech Connect (OSTI)

The In-Tank Precipitation process (ITP) receives High Level Waste (HLW) supernatant liquid containing radionuclides in waste processing tank 48H. Sodium tetraphenylborate, NaTPB, and monosodium titanate (MST), NaTi{sub 2}O{sub 5}H, are added for removal of radioactive Cs and Sr, respectively. In addition to removal of radio-strontium, MST will also remove plutonium and uranium. The majority of the feed solutions to ITP will come from the dissolution of supernate that had been concentrated by evaporation to a crystallized salt form, commonly referred to as saltcake. The concern for criticality safety arises from the adsorption of U and Pt onto MST. If sufficient mass and optimum conditions are achieved then criticality is credible. The concentration of u and Pt from solution into the smaller volume of precipitate represents a concern for criticality. This report supplements WSRC-TR-93-171, Nuclear Criticality Safety Bounding Analysis For The In-Tank-Precipitation (ITP) Process. Criticality safety in ITP can be analyzed by two bounding conditions: (1) the minimum safe ratio of MST to fissionable material and (2) the maximum fissionable material adsorption capacity of the MST. Calculations have provided the first bounding condition and experimental analysis has established the second. This report combines these conditions with canyon facility data to evaluate the potential for criticality in the ITP process due to the adsorption of the fissionable material from solution. In addition, this report analyzes the potential impact of increased U loading onto MST. Results of this analysis demonstrate a greater safety margin for ITP operations than the previous analysis. This report further demonstrates that the potential for criticality in the ITP process due to adsorption of fissionable material by MST is not credible.

Bess, C.E.

1994-04-22T23:59:59.000Z

331

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

E-Print Network [OSTI]

that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement (LOCA) and loss of flow accident (LOFA); Inventories; Safety; Environment studies 1. Background

332

Analysis of improvements in system efficiency and safety at highway-railroad-pedestrian grade crossings  

E-Print Network [OSTI]

The purpose of this project was to perform micro-simulation analyses on intersections near Highway-Railroad Grade Crossings to determine if controlling mean train speed and train speed variability would improve safety and reduce delays. The first...

Tydlacka, Jonathan Michael

2013-02-22T23:59:59.000Z

333

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Waiver for the Use of Modified Primary Containment Vessel (PCV)  

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) staff, at the request of the U.S. Department of Energy (DOE), on the Waiver for the Use of Modified Primary Containment Vessels (PCV). The waiver is to be used to support a limited number of shipments of fuel for the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) Project in support of the National Aeronautics and Space Administration's (NASA's) Mars Science Laboratory (MSL) mission. Under the waiver, an inventory of existing national security PCVs will be converted to standard PCVs. Both types of PCVs are currently approved for use by the Office of Nuclear Energy. LLNL has previously reviewed the national security PCVs under Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision c, dated June 2007 (Addendum 1). The safety analysis of the package is documented in the Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package (i.e., the Mound 1KW SARP, or the SARP) where the standard PCVs have been reviewed by LLNL. The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The waiver requests an exemption, claiming safety equivalent to the requirements specified in 10 CFR 71.12, Specific Exemptions, and will lead to a letter amendment to the CoC. Under the waiver, the Office of Radioisotope Power Systems, NE-34, is seeking an exemption from 10 CFR 71.19(d)(1), Previously Approved Package,[5] which states: '(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in {section}71.71 and 71.73.' The LLNL staff had previously reviewed a request from Idaho National Laboratory (INL) to reconfigure national security PCVs to standard PCVs. With a nominal 50% reduction in both the height and the volume, the LLNL staff initially deemed the modifications to be significant, which would not be allowed under the provisions of 10 CFR 71.19(d)(1)--see above. As a follow-up, the DOE requested additional clarification from the Nuclear Regulatory Commission (NRC). The NRC concluded that the reconfiguration would be a new fabrication, and that an exemption to the regulations would be required to allow its use, as per the requirements specified in 10 CFR 71.19(c)(1), Previously Approved Package: '(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation '-85' in the identification number of the NRC CoC, may be used under the general license of {section}71.17 with the following additional conditions: (1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with 71.85(c).' Although the preferred approach toward the resolution of this issue would be for the applicant to submit an updated SARP, the applicant has stated that the process of updating the Model Mound 1KW Package SARP is a work that is in progress, but that the updated SARP is not yet ready for submittal. The applicant has to provide a submittal, proving that the package meets the '-96' requirements of International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1, in order to fabricate approved packagings after December 31, 2006. The applicant has further stated that all other packaging features, as described in the currently approved Model Mound 1KW Package SARP, remain unchanged. This report documents the LLNL review of the waiver request. The specific review for each SARP Chapter is documented.

West, M; Hafner, R

2008-05-05T23:59:59.000Z

334

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

Broader source: Energy.gov [DOE]

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

335

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

SciTech Connect (OSTI)

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

CARRELL, R.D.

2002-07-16T23:59:59.000Z

336

Technical safety requirements for the Auxiliary Hot Cell Facility (AHCF).  

SciTech Connect (OSTI)

These Technical Safety Requirements (TSRs) identify the operational conditions, boundaries, and administrative controls for the safe operation of the Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, in compliance with 10 CFR 830, 'Nuclear Safety Management.' The bases for the TSRs are established in the AHCF Documented Safety Analysis (DSA), which was issued in compliance with 10 CFR 830, Subpart B, 'Safety Basis Requirements.' The AHCF Limiting Conditions of Operation (LCOs) apply only to the ventilation system, the high efficiency particulate air (HEPA) filters, and the inventory. Surveillance Requirements (SRs) apply to the ventilation system, HEPA filters, and associated monitoring equipment; to certain passive design features; and to the inventory. No Safety Limits are necessary, because the AHCF is a Hazard Category 3 nuclear facility.

Seylar, Roland F.

2004-02-01T23:59:59.000Z

337

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

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

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

1998-05-01T23:59:59.000Z

338

Monitored Retrievable Storage System Requirements Document. Revision 1  

SciTech Connect (OSTI)

This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design.

Not Available

1994-03-01T23:59:59.000Z

339

CRAD, Facility Safety- Unreviewed Safety Question Requirements  

Broader source: Energy.gov [DOE]

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

340

CRAD, Facility Safety- Technical Safety Requirements  

Broader source: Energy.gov [DOE]

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

Note: This page contains sample records for the topic "documented 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

CRAD, Facility Safety- Nuclear Facility Safety Basis  

Broader source: Energy.gov [DOE]

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

342

MIDAS documentation  

SciTech Connect (OSTI)

MIDAS, the Master Information and Data Acquisition System, is a computerized system being implemented on FFTF to control the work released to the plant. The purpose of this system is to demonstrate the safety enhancement provided for LMFBRs when the operator has instant recall to the status of all work released to the plant, the interrelationships between functional equipment groups in the plant, and the relationships of equipment to safety functions.

Seeman, S.E.

1981-09-01T23:59:59.000Z

343

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

Broader source: Energy.gov [DOE]

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

344

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

345

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

346

Prism++ Documentation University of Erlangen  

E-Print Network [OSTI]

! "monk-it" Prism++ Documentation University of Erlangen Computer Networks and Communication.2.1. Correlation engine management . . . . . . . . . . . . . . . . . . . . 8 2.2.2. Analysis of collected events

Breu, Ruth

347

High-level waste tank farm set point document  

SciTech Connect (OSTI)

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

Anthony, J.A. III

1995-01-15T23:59:59.000Z

348

Lawrence Livermore National Laborotory Safety Basis Assessment...  

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

Lawrence Livermore National Laboratory Safety Basis Assessment INTRODUCTION This site visit report documents the collective results of the review of Lawrence Livermore National...

349

Chemical Safety Program - Library | Department of Energy  

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

Library Chemical Safety Program - Library Operating Experience Level 3 OSHA's Revised Hazard Communication Standard Safe Management of Mercury DOE Documents & Guidelines Chemical...

350

Integrated Safety Management Workshop Registration, PIA, Idaho...  

Office of Environmental Management (EM)

Safety Management Workshop Registration, PIA, Idaho National Laboratory More Documents & Publications TRAIN-PIA.pdf Occupational Medicine - Assistant PIA, Idaho National Laboratory...

351

Regulatory guidance document  

SciTech Connect (OSTI)

The Office of Civilian Radioactive Waste Management (OCRWM) Program Management System Manual requires preparation of the OCRWM Regulatory Guidance Document (RGD) that addresses licensing, environmental compliance, and safety and health compliance. The document provides: regulatory compliance policy; guidance to OCRWM organizational elements to ensure a consistent approach when complying with regulatory requirements; strategies to achieve policy objectives; organizational responsibilities for regulatory compliance; guidance with regard to Program compliance oversight; and guidance on the contents of a project-level Regulatory Compliance Plan. The scope of the RGD includes site suitability evaluation, licensing, environmental compliance, and safety and health compliance, in accordance with the direction provided by Section 4.6.3 of the PMS Manual. Site suitability evaluation and regulatory compliance during site characterization are significant activities, particularly with regard to the YW MSA. OCRWM`s evaluation of whether the Yucca Mountain site is suitable for repository development must precede its submittal of a license application to the Nuclear Regulatory Commission (NRC). Accordingly, site suitability evaluation is discussed in Chapter 4, and the general statements of policy regarding site suitability evaluation are discussed in Section 2.1. Although much of the data and analyses may initially be similar, the licensing process is discussed separately in Chapter 5. Environmental compliance is discussed in Chapter 6. Safety and Health compliance is discussed in Chapter 7.

NONE

1994-05-01T23:59:59.000Z

352

Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants  

SciTech Connect (OSTI)

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

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

2007-08-01T23:59:59.000Z

353

Fluor Daniel Hanford contract standards/requirements identification document  

SciTech Connect (OSTI)

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

Bennett, G.L.

1997-04-24T23:59:59.000Z

354

Model-Based Quantitative Safety Analysis of Matlab Simulink / Stateflow Models  

E-Print Network [OSTI]

ISO 26262 [Int11] for the automotive domain or DO-178C [sta12] for the avionics domain recommend is widely used to design systems in the automotive and avionics domains. ISO 26262 and DO-178C require (FMEA) [Int91] or Fault Tree Anal- ysis (FTA) [U.S81]. This gives rise to the question how safety

Leue, Stefan

355

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

E-Print Network [OSTI]

Trees and Failure Modes and Effects Analyses (FMEAs) can be automatically derived from a topological of failure patterns in conjunction with automated fault tree and FMEA synthesis algorithms can help on safety is universal, the issue is perhaps more urgent in the automotive industry which currently

Boyer, Edmond

356

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor Technology Subcommittee ofJanuary

357

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor Technology Subcommittee

358

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor Technology SubcommitteeOctober 13,

359

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor Technology SubcommitteeOctober

360

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor Technology

Note: This page contains sample records for the topic "documented 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

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor TechnologyAugust 18, 1999

362

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor TechnologyAugust 18, 1999July 20,

363

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor TechnologyAugust 18, 1999July

364

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014Nuclear

365

WIPP Documents - Quality Assurance and Safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert SouthwestTechnologies | Blandine Jerome VelenciaNewsNewsFederalQuality Assurance

366

Independent Oversight Targeted Review of the Safety Significant...  

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

Site Office SAR Safety Analysis Report SC Safety Class SCI SuspectCounterfeit Items SME Subject Matter Expert SMP Safety Management Program SR Surveillance Requirement SS...

367

Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II Safety Program  

SciTech Connect (OSTI)

Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated.

Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutoy, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N. [RRC Kurchatov Institute, Moscow (Russian Federation)] [and others

1994-11-01T23:59:59.000Z

368

Application of System-Theoretic Process Analysis to Engineered Safety Features-Component Control System  

E-Print Network [OSTI]

(FTA) or failure mode and effects analysis (FMEA), have been extensively used for decades. However Modes and Effects analysis (FMEA), Hazards and Operability Analysis (HAZOP), etc., have been proposed

369

Challenges in the Modeling and Quantitative Analysis of Safety-Critical Automotive Systems!  

E-Print Network [OSTI]

! Probabilistic FMEA! Probabilistic Analysis of System Architectures! ! Conclusion! 3! #12;ISO 26262: Road! ,,identify Failures"! - Qualitative FMEA! ! - Qualitative Fault Tree Analysis! ! - Event Tree Analysis! Quantitative Methods! ,,predict frequency of failures"! - Quantitative FMEA! ! - Quantitative Fault Tree

Leue, Stefan

370

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

SciTech Connect (OSTI)

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

371

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

372

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

373

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

E-Print Network [OSTI]

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

Galvez, Cristhian

2011-01-01T23:59:59.000Z

374

Safety of magnetic fusion facilities: Requirements  

SciTech Connect (OSTI)

This Standard identifies safety requirements for magnetic fusion facilities. Safety functions are used to define outcomes that must be achieved to ensure that exposures to radiation, hazardous materials, or other hazards are maintained within acceptable limits. Requirements applicable to magnetic fusion facilities have been derived from Federal law, policy, and other documents. In addition to specific safety requirements, broad direction is given in the form of safety principles that are to be implemented and within which safety can be achieved.

NONE

1996-05-01T23:59:59.000Z

375

Analysis of consequences of postulated solvent fires in Hanford site waste tanks  

SciTech Connect (OSTI)

This document contains the calculations that support the accident analyses for accidents involving organic solvents. This work was performed to support the Basis for Interim Operation (BIO) and the Final Safety Analysis Report (FSAR) for Tank Waste Remediation Systems (TWRS).

Cowley, W.L., Westinghouse Hanford

1996-08-12T23:59:59.000Z

376

Ferrocyanide Safety Program rationale for removing six tanks from the safety watch list  

SciTech Connect (OSTI)

This report documents an in-depth study of single-shell tanks containing ferrocyanide wastes. Topics include: safety assessments, tank histories, supportive documentation about interim stabilization and planned remedial activities.

Borsheim, G.L.

1993-09-01T23:59:59.000Z

377

Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1  

SciTech Connect (OSTI)

This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.

NONE

1995-01-18T23:59:59.000Z

378

FAQS Qualification Card – Criticality Safety  

Broader source: Energy.gov [DOE]

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).

379

Environmental Health and Safety Assessment  

E-Print Network [OSTI]

Environmental Health and Safety Assessment Program Manual 7/15/2013 #12;Environmental Health/26/2013. The most recent version of this document is available electronically at: http://sp.ehs.cornell.edu/env/general-environmental-management/environmental.........................................................................................................................4 #12;Environmental Health and Safety Assessment Program Manual Approved by: (Barb English) Last

Pawlowski, Wojtek

380

Electrical Equipment Inspection Program Electrical Safety  

E-Print Network [OSTI]

Electrical Equipment Inspection Program Electrical Safety SLAC-I-730-0A11A-001-R003 23 March 2005 Document Title: Electrical Equipment Inspection Program Original Publication Date: 19 January 2005 Revised Publication Date: 23 March 2005 (updated 29 November 2010) Department: Electrical Safety Document Number: SLAC

Wechsler, Risa H.

Note: This page contains sample records for the topic "documented 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.


381

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

382

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

383

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

384

Documenting Spreadsheets  

E-Print Network [OSTI]

This paper discusses spreadsheets documentation and new means to achieve this end by using Excel's built-in "Comment" function. By structuring comments, they can be used as an essential tool to fully explain spreadsheet. This will greatly facilitate spreadsheet change control, risk management and auditing. It will fill a crucial gap in corporate governance by adding essential information that can be managed in order to satisfy internal controls and accountability standards.

Payette, Raymond

2008-01-01T23:59:59.000Z

385

Document13  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to UserProduct:Directives Templates The

386

Environmental Documents  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series toESnet4: NetworkingEnvironment Environment

387

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

388

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

389

SAFETY ANALYSIS OF A RADIO-BASED CROSSING CONTROL SYSTEM USING FORMAL  

E-Print Network [OSTI]

), failure mode and e#11;ects anal- ysis (FMEA) (Reifer 1979), or hazard and oper- ability analysis (HAZOP

Reif, Wolfgang

390

An OSHA based approach to safety analysis for nonradiological hazardous materials  

SciTech Connect (OSTI)

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

Yurconic, M.

1992-08-01T23:59:59.000Z

391

An OSHA based approach to safety analysis for nonradiological hazardous materials  

SciTech Connect (OSTI)

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

Yurconic, M.

1992-08-01T23:59:59.000Z

392

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

SciTech Connect (OSTI)

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

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

1999-12-01T23:59:59.000Z

393

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 (OSTI)

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

394

Environment, Health & Safety Division 31 July, 2009  

E-Print Network [OSTI]

Environment, Health & Safety Division 31 July, 2009 MEMORANDUM To: Division Safety Coordinators Division Liaisons All JHA Users From: John Seabury Environment, Health & Safety Division Subject: Job Hazard Analysis ­ Description of Work Rev 5 Discussion, Requirements, Helpful Hints and Examples

395

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

396

Thermal reactor safety  

SciTech Connect (OSTI)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

Not Available

1980-06-01T23:59:59.000Z

397

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

SciTech Connect (OSTI)

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

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

1992-07-01T23:59:59.000Z

398

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

SciTech Connect (OSTI)

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

Not Available

1991-01-01T23:59:59.000Z

399

INL Procurement Document and Forms  

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

Excavation Checklist Word 432.58 Job Safety Analysis PRD-1007 Word 432.65 Construction LockoutTagout Checklist Word 432.A31 Competent Person Roster Word 432.A42 Manpower and...

400

Document Information  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed Newcatalyst phasesDataTranslocationDiurnal Cycle ofDo New55 FederalChapter5 Document

Note: This page contains sample records for the topic "documented 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

Document Delivery  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to UserProduct:Directives Templates The OfficeDitchDivya401

402

Reference Documents  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared at 278, 298,NIST 800-53 NationalTreatment.Reference-Documents Sign

403

Reference Documents  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared at 278, 298,NIST 800-53 NationalTreatment.Reference-Documents

404

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

SciTech Connect (OSTI)

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 10{sup 6} cm{sup ?1}) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 10{sup 6} cm{sup ?1}) the limits were exceeded.

Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Dewayatna, Winter, E-mail: winter@batan.go.id [Center for Nuclear Fuel Technology, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Yos, E-mail: soj@batan.go.id [Center for Nuclear Equipment and Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

2014-09-30T23:59:59.000Z

405

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

406

Canister Storage Building (CSB) Hazard Analysis Report  

SciTech Connect (OSTI)

This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other controls required to protect the public, workers, and environment.

POWERS, T.B.

2000-03-16T23:59:59.000Z

407

Safety and Security Interface Technology Initiative  

SciTech Connect (OSTI)

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

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

2007-05-01T23:59:59.000Z

408

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

SciTech Connect (OSTI)

The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The National Nuclear Security Administration (NNSA) submitted SAND Report SAND2009-5822 to NRC that documented the incorporation of plutonium (Pu) metal as a new payload for the PAT-1 package. NRC responded with a Request for Additional Information (RAI), identifying information needed in connection with its review of the application. The purpose of this SAND report is to provide the authors responses to each RAI. SAND Report SAND2010-6106 containing the proposed changes to the Addendum is provided separately.

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

2010-09-01T23:59:59.000Z

409

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

SciTech Connect (OSTI)

The U.S. Nuclear Regulatory Commission on July 27, 2001 approved Revision 19 of the TRUPACT-II Safety Analysis Report (SAR) and the associated TRUPACT-II Authorized Methods for Payload Control (TRAMPAC). Key initiatives in Revision 19 included matrix depletion, unlimited mixing of shipping categories, a flammability assessment methodology, and an alternative methodology for the determination of flammable gas generation rates. All U.S. Department of Energy (DOE) sites shipping transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) were required to implement Revision 19 methodology into their characterization and waste transportation programs by May 20, 2002. An implementation process was demonstrated by the Rocky Flats Environmental Technology Site (RFETS) in Golden, Colorado. The three-part process used by RFETS included revision of the site-specific TRAMPAC, an evaluation of the contact-handled TRU waste inventory against the regulations in Revision 19, and design and development of software to facilitate future inventory analyses.

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

2003-02-25T23:59:59.000Z

410

Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II safety program  

SciTech Connect (OSTI)

Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}

Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutov, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.; Chunyaev, E.I. [RRC Kurchatov Institute, Moscow 123182 (Russian Federation); Marshall, A.C. [International Nuclear Safety, Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States); Sapir, J.L.; Pelowitz, D.B. [Reactor Design and Analysis Group, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

1995-01-20T23:59:59.000Z

411

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

412

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

SciTech Connect (OSTI)

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

413

Order Module--NNSA Orders Self-Study Program Safety Basis Documentatio...  

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

NNSA Orders Self-Study Program Safety Basis Documentation Order Module--NNSA Orders Self-Study Program Safety Basis Documentation The familiar level of this module is divided into...

414

Preservation of FFTF Data Related to Passive Safety Testing  

SciTech Connect (OSTI)

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 1980’s. 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

415

Lift truck safety review  

SciTech Connect (OSTI)

This report presents safety information about powered industrial trucks. The basic lift truck, the counterbalanced sit down rider truck, is the primary focus of the report. Lift truck engineering is briefly described, then a hazard analysis is performed on the lift truck. Case histories and accident statistics are also given. Rules and regulations about lift trucks, such as the US Occupational Safety an Health Administration laws and the Underwriter`s Laboratories standards, are discussed. Safety issues with lift trucks are reviewed, and lift truck safety and reliability are discussed. Some quantitative reliability values are given.

Cadwallader, L.C.

1997-03-01T23:59:59.000Z

416

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

Broader source: Energy.gov [DOE]

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

417

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

Broader source: Energy.gov [DOE]

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

418

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

Broader source: Energy.gov [DOE]

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

419

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

Broader source: Energy.gov [DOE]

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

420

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

Broader source: Energy.gov [DOE]

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

Note: This page contains sample records for the topic "documented 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

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

Broader source: Energy.gov [DOE]

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

422

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

Broader source: Energy.gov [DOE]

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

423

Facility Safety  

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

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

424

Criticality Safety Controls Implementation Inspection Criteria, Approach, and Lines of Inquiry, October 23, 2009, (HSS CRAD 64-18, Rev 0)  

Broader source: Energy.gov [DOE]

DOE has set expectations for implementing criticality safety controls that are selected to provide preventive and/or mitigative functions for specific potential accident scenarios. There are additional expectations for criticality safety controls that are also designated as Specific Administrative Controls (SACs) (see HSS CRAD 64-32). Also, in instances when the review addresses functionality and operability of structures, systems, and components (SSCs) of nuclear facilities specifically required for criticality safety per the facility's documented safety analysis (DSA), see HSS CRAD 64-11.

425

QUEST2: Release 1, SA/Release 1 supporting documents deliverable set  

SciTech Connect (OSTI)

This document contains deliverables which reflect the last of the System Architecture phase analysis for the Quality, Environmental, Safety Tracking System redesign (QUEST2) project. These deliverables are focused on the final insights required to start functional design of the first QUEST2 release. They include the data definitions, conversion rules, standards for design and user interface, performance criteria, and rules to be followed during the prototyping activity described in the Project Management Plan.

Braaten, F.D.

1995-02-27T23:59:59.000Z

426

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

SciTech Connect (OSTI)

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

427

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

SciTech Connect (OSTI)

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

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

1992-07-01T23:59:59.000Z

428

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

SciTech Connect (OSTI)

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

429

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

430

Operating Experience Level 3, Safety Concern: Roll-up Doors ...  

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

Level 3, Safety Concern: Roll-up Doors Operating Experience Level 3, Safety Concern: Roll-up Doors August 14, 2014 This Operating Experience Level 3 (OE-3) document provides...

431

Nuclear Safety Policy - DOE Directives, Delegations, and Requirements  

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

ARCHIVED SEN-35-91, Nuclear Safety Policy by jnichols Functional areas: Environment, Safety, and Health, Canceled by DOE P 420.1 n3591.pdf -- PDF Document, 20 KB Writer: jnichols...

432

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

SciTech Connect (OSTI)

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

Not Available

1990-12-01T23:59:59.000Z

433

Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation  

SciTech Connect (OSTI)

The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

Chiang, R. T. [AREVA Inc., 303 Ravendale Drive, Mountain View, CA 94043 (United States)

2013-07-01T23:59:59.000Z

434

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

SciTech Connect (OSTI)

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

435

XML Document XML Document Types and Validation  

E-Print Network [OSTI]

XML Document Types and Validation IIM-I340 Objectives XML Document Types and Validation IIM-I340 April, 2010 #12;XML Document Types and Validation IIM-I340 Objectives Learning Objectives Understand: The need for validation Two ways to specify validity: Document Type Definitions (DTDs) XML Schemas #12;XML

Weber, Gregory D.

436

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

SciTech Connect (OSTI)

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

437

Omega documentation  

SciTech Connect (OSTI)

OMEGA is a CRAY I computer program that controls nine codes used by LLNL Physical Data Group for: 1) updating the libraries of evaluated data maintained by the group (UPDATE); 2) calculating average values of energy deposited in secondary particles and residual nuclei (ENDEP); 3) checking the libraries for internal consistency, especially for energy conservation (GAMCHK); 4) producing listings, indexes and plots of the library data (UTILITY); 5) producing calculational constants such as group averaged cross sections and transfer matrices for diffusion and Sn transport codes (CLYDE); 6) producing and updating standard files of the calculational constants used by LLNL Sn and diffusion transport codes (NDFL); 7) producing calculational constants for Monte Carlo transport codes that use group-averaged cross sections and continuous energy for particles (CTART); 8) producing and updating standard files used by the LLNL Monte Carlo transport codes (TRTL); and 9) producing standard files used by the LANL pointwise Monte Carlo transport code MCNP (MCPOINT). The first four of these functions and codes deal with the libraries of evaluated data and the last five with various aspects of producing calculational constants for use by transport codes. In 1970 a series, called PD memos, of internal and informal memoranda was begun. These were intended to be circulated among the group for comment and then to provide documentation for later reference whenever questions arose about the subject matter of the memos. They have served this purpose and now will be drawn upon as source material for this more comprehensive report that deals with most of the matters covered in those memos.

Howerton, R.J.; Dye, R.E.; Giles, P.C.; Kimlinger, J.R.; Perkins, S.T.; Plechaty, E.F.

1983-08-01T23:59:59.000Z

438

Experiment Design and Analysis Guide - Neutronics & Physics  

SciTech Connect (OSTI)

The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

Misti A Lillo

2014-06-01T23:59:59.000Z

439

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

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

Work Planning and Control Guidance Document May 15, 2013 Presenters: Steele Coddington, NSTec, Las Vegas, and John McDonald, WRPS, Hanford Topics Covered: Integration of Safety...

440

Safety Planning Guidance for Hydrogen and Fuel Cell Projects  

Fuel Cell Technologies Publication and Product Library (EERE)

This guidance document provides information on safety requirements for hydrogen and fuel cell projects funded by the U.S. Department of Energy Fuel Cell Technologies Program.

Note: This page contains sample records for the topic "documented 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

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

E-Print Network [OSTI]

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

Wenzel, Thomas P.

2010-01-01T23:59:59.000Z

442

Preliminary nuclear safety assessment of the NEPST (Topaz II) space reactor program  

SciTech Connect (OSTI)

The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

Marshall, A.C.

1993-01-01T23:59:59.000Z

443

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  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana. DOCUMENTS AVAILABLEReport 2009SiteMajor Maintenance at Philpott Lake

444

Tag: Safety  

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

8all en Best Practices Workshop for Safety Culture http:www.y12.doe.goveshbest-practices-workshop-safety-culture

445

WHC fire hazards analysis policy  

SciTech Connect (OSTI)

The purpose of this document is to establish the fire protection policy for Westinghouse Hanford Company (WHC) relative to US Department of Energy (DOE) directives for Fire Hazards Analyses (FHAs) and their relationship to facility Safety Analysis Reports (SARs) as promulgated by the DOE Richland Operations Office.

Evans, C.B.

1994-04-01T23:59:59.000Z

446

Safety Information for Families  

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

447

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

SciTech Connect (OSTI)

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

TOMASZEWSKI, T.A.

2000-04-25T23:59:59.000Z

448

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

SciTech Connect (OSTI)

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the cos

Enercon Services, Inc.

2011-03-14T23:59:59.000Z

449

Biological Safety  

Broader source: Energy.gov [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.

450

Facility Safety  

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

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

451

Facility Safety  

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

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

2012-12-04T23:59:59.000Z

452

Facility Safety  

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

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

453

Facility Safety  

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

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

454

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

SciTech Connect (OSTI)

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

455

Advanced Modeling of Multicomponent Vaporization/Condensation Phenomena for a Reactor Safety Analysis Code SIMMER-III  

SciTech Connect (OSTI)

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis code SIMMER-III, which analyzes relatively short-time-scale phenomena relevant to accident sequences of CDAs. The model characterizes the V/C process associated with phase transition through heat-transfer and mass-diffusion limited models to follow the time evolution of the rector core under CDA conditions. The heat-transfer limited model describes the nonequilibrium phase-transition processes occurring at interfaces, while the mass-diffusion limited model is employed to represent effects of noncondensable gases and multicomponent mixture on V/C processes. Verification of the model and method employed in the multicomponent V/C model of SIMMER-III was performed successfully by analyzing two series of condensation experiments. (authors)

Koji Morita; Tatsuya Matsumoto; Ryo Akasaka; Kenji Fukuda [Kyushu University, 6-10-1, Hakozaki, Higashi-ku, Fukuoka 812-81 (Japan); Tohru Suzuki; Yoshiharu Tobita; Hidemasa Yamano; Satoru Kondo [Japan Nuclear Cycle Development Institute 4002, Narita, O-arai, Ibaraki 311-1393 (Japan)

2002-07-01T23:59:59.000Z

456

Fuzzy Typing for Document Management Alison HUETTNER  

E-Print Network [OSTI]

Fuzzy Typing for Document Management Alison HUETTNER Clairvoyance Corporation 5301 Fifth Avenue method of document analysis and management, based on a combination of techniques from NLP and fuzzy logic typing for document management. The fuzzy typing approach is general in scope and can be applied to many

Wu, Dekai

457

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.  

SciTech Connect (OSTI)

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-03-01T23:59:59.000Z

458

Safety evaluation of a hydrogen fueled transit bus  

SciTech Connect (OSTI)

Hydrogen fueled vehicle demonstration projects must satisfy management and regulator safety expectations. This is often accomplished using hazard and safety analyses. Such an analysis has been completed to evaluate the safety of the H2Fuel bus to be operated in Augusta, Georgia. The evaluation methods and criteria used reflect the Department of Energy`s graded approach for qualifying and documenting nuclear and chemical facility safety. The work focused on the storage and distribution of hydrogen as the bus motor fuel with emphases on the technical and operational aspects of using metal hydride beds to store hydrogen. The safety evaluation demonstrated that the operation of the H2Fuel bus represents a moderate risk. This is the same risk level determined for operation of conventionally powered transit buses in the United States. By the same criteria, private passenger automobile travel in the United States is considered a high risk. The evaluation also identified several design and operational modifications that resulted in improved safety, operability, and reliability. The hazard assessment methodology used in this project has widespread applicability to other innovative operations and systems, and the techniques can serve as a template for other similar projects.

Coutts, D.A.; Thomas, J.K.; Hovis, G.L.; Wu, T.T. [Westinghouse Savannah River Co., Aiken, SC (United States)

1997-12-31T23:59:59.000Z

459

Accident Analysis Guidance for Completion of 10 CFR 830-Compliant DSAs  

SciTech Connect (OSTI)

Safety analysis contractors responsible for existing nuclear facilities are required to submit a Documented Safety Analysis to the Department of Energy for approval by April 2003. Recognizing that schedule and resource limitations may be significant, an initiative is underway to make available a set of guidance tools. The guidance is in the form of a peer-reviewed Accident Analysis Guidebook, a series of application guides for ''safe harbor'' computer codes, establishment of a configuration-controlled collection of safety analysis software and a central registry to maintain it, and periodic analytical training on accident analysis methods. Delivery of the majority of these products is scheduled to be in FY 2003.

Vincent, A.

2002-06-24T23:59:59.000Z

460

INTEGRATED SAFETY MANAGEMENT SYSTEM SAFETY CULTURE IMPROVEMENT INITIATIVE  

SciTech Connect (OSTI)

In 2007, the Department of Energy (DOE) identified safety culture as one of their top Integrated Safety Management System (ISMS) related priorities. A team was formed to address this issue. The team identified a consensus set of safety culture principles, along with implementation practices that could be used by DOE, NNSA, and their contractors. Documented improvement tools were identified and communicated to contractors participating in a year long pilot project. After a year, lessons learned will be collected and a path forward determined. The goal of this effort was to achieve improved safety and mission performance through ISMS continuous improvement. The focus of ISMS improvement was safety culture improvement building on operating experience from similar industries such as the domestic and international commercial nuclear and chemical industry.

MCDONALD JA JR

2009-01-16T23:59:59.000Z

Note: This page contains sample records for the topic "documented 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

Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core  

SciTech Connect (OSTI)

Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

PARMA JR.,EDWARD J.

2000-01-01T23:59:59.000Z

462

ENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION  

E-Print Network [OSTI]

: FS Vancouver: Ops CHEMICAL SAFETY 27265 CONTRACTOR SAFETY 23867 EARLY RETURN TO WORK 23011 EMERGENCYENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION SIMON FRASER UNIVERSITY ENVIRONMENTAL HEALTH & SAFETY DEPARTMENT Discovery Park - MTF 8888 University Drive Burnaby, British Columbia Canada V5

463

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

SciTech Connect (OSTI)

The orbiter and probe portions of the NASA Galileo spacecraft contain components which require auxiliary heat during the mission. To meet these needs, the Department of Energy's (DOE's) Office of Special Nuclear Projects (OSNP) has sponsored the design, fabrication, and testing of a one-watt encapsulated plutonium dioxide-fueled thermal heater named the Light-Weight Radioisotope Heater Unit (LWRHU). This report addresses the radiological risks which might be encountered by people both at the launch area and worldwide should postulate mission failures or malfunctions occur, which would result in the release of the LWRHUs to the environment. Included are data from the design, mission descriptions, postulated accidents with their consequences, test data, and the derived source terms and personnel exposures for the various events.

Johnson, E.W.

1985-10-01T23:59:59.000Z

464

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy(National1 -OSSGasofOfficeKathleen Hanford Site

465

General safety basis development guidance for environmental restoration decontamination and decommissioning  

SciTech Connect (OSTI)

Safety analyses have the objective of contributing to two essential ingredients of a successful operation. The first is promoting the safety of the operation through worker involvement in information development (safety basis). The second is obtaining approval to conduct the operation (authorization). Typically these ingredients are assembled under separate programs covered by separate DOE requirements. DOE authorization relies on successful development of a document containing up to 21 topics written in terms and language suited to reviewers and approvers. Safety relies on successful training and procedures that convert the technical documented information into terms and language understandable to the worker. This separation can lead to successful incorporation of one ingredient independent of the other. At best, this separation may result in a safe but unauthorized operation; at worst, the separation may result in an unsafe operation authorized to proceed. This guide is based on experiences gained by contractors who have integrated rather than separated the safety and authorization. The short duration of ER/D&D activities, the uncertainties of hazards, and the publicly expressed desire for demonstrable progress in cleanup activities add emphasis to the need to integrate rather than separate and develop new programs. Experience-based information has been useful to workers, safety analysis practitioners, and reviewers in the following ways: (1) Acquiring or developing the needed information in a useful form; (2) Managing the uncertainties during activity development and operation; (3) Identifying the subset of applicable requirements for an activity; (4) Developing the appropriate level of documentation detail for a specific activity; and (5) Increasing the usefulness and use of safety analysis (ownership).

Ellingson, D.R.; Kerr, N. [Westinghouse Hanford Co., Richland, WA (United States); Bohlander, K. [EG and G Rocky Flats, Golden, CO (United States); Hansen, J. [Westinghouse Savannah River Co., Aiken, SC (United States); Crowley, W. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)

1994-02-01T23:59:59.000Z

466

Occupational Safety Review of High Technology Facilities  

SciTech Connect (OSTI)

This report contains reviews of operating experiences, selected accident events, and industrial safety performance indicators that document the performance of the major US DOE magnetic fusion experiments and particle accelerators. These data are useful to form a basis for the occupational safety level at matured research facilities with known sets of safety rules and regulations. Some of the issues discussed are radiation safety, electromagnetic energy exposure events, and some of the more widespread issues of working at height, equipment fires, confined space work, electrical work, and other industrial hazards. Nuclear power plant industrial safety data are also included for comparison.

Lee Cadwallader

2005-01-31T23:59:59.000Z

467

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

Broader source: Energy.gov [DOE]

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

468

Research Data Management - Documentation  

E-Print Network [OSTI]

This short (16 slide) presentation "Documentation" explains the need to create documentation using several clear examples of the benefits from good documentation (and what can happen if it is not good), making research reproducible...

Collins, Anna

2013-01-01T23:59:59.000Z

469

Hazard Analysis Database Report  

SciTech Connect (OSTI)

The Hazard Analysis Database was developed in conjunction with the hazard analysis activities conducted in accordance with DOE-STD-3009-94, Preparation Guide for U S . Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, for HNF-SD-WM-SAR-067, Tank Farms Final Safety Analysis Report (FSAR). The FSAR is part of the approved Authorization Basis (AB) for the River Protection Project (RPP). This document describes, identifies, and defines the contents and structure of the Tank Farms FSAR Hazard Analysis Database and documents the configuration control changes made to the database. The Hazard Analysis Database contains the collection of information generated during the initial hazard evaluations and the subsequent hazard and accident analysis activities. The Hazard Analysis Database supports the preparation of Chapters 3 ,4 , and 5 of the Tank Farms FSAR and the Unreviewed Safety Question (USQ) process and consists of two major, interrelated data sets: (1) Hazard Analysis Database: Data from the results of the hazard evaluations, and (2) Hazard Topography Database: Data from the system familiarization and hazard identification.

GRAMS, W.H.

2000-12-28T23:59:59.000Z

470

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

SciTech Connect (OSTI)

This document, the Standards/Requirements Identification Document (S/RIDS) for the subject facility, represents the necessary and sufficient requirements to provide an adequate level of protecti