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1

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

2

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

3

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

4

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

5

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

6

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

7

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

8

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

9

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

10

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

11

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

12

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

13

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

14

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

15

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

16

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

17

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...

18

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

19

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...

20

Ferrocyanide safety project ferrocyanide aging studies. Final report  

SciTech Connect (OSTI)

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

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

1996-06-01T23:59:59.000Z

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


21

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...

22

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

SciTech Connect (OSTI)

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

T. J. Rodovsky

2006-03-01T23:59:59.000Z

23

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.

24

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...

25

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

26

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

27

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

SciTech Connect (OSTI)

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

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

1994-02-01T23:59:59.000Z

28

Final Report - Hydrogen Delivery Infrastructure Options Analysis...  

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

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

29

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

30

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

31

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

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

13-17, 2010. The assessment revealed that LSO has implemented appropriate plans, procedures, and mechanisms to oversee implementation of the safety basis and unreviewed safety...

32

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

33

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...

34

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...

35

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

36

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

37

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.

38

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

39

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

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

* Control of chemical attack * Summary 2 * Summary Modular HTGR Safety Design Philosophy Top-level Requirement * Worker doses within 20% Design Solution * Control...

40

Microsoft Word - Technical Safety and Licensing Issues _Final...  

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

event is neither necessary nor desirable in implementing the safety design philosophy of the NGNP and modular HTRs. The strategy to license this approach with the...

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


41

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

42

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

43

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

44

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

E-Print Network [OSTI]

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

Ahmad, Sajjad

45

DOE high-level waste tank safety program. Final report  

SciTech Connect (OSTI)

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

NONE

1998-11-01T23:59:59.000Z

46

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...

47

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

E-Print Network [OSTI]

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

Farritor, Shane

48

Enrollment Analysis Final for Fall 2014  

E-Print Network [OSTI]

) November 6, 2014 Project Request: Enrollment Analysis ­ Final for Fall 2014. Requested by: Dr. Brooks Keel, President; Dr. Teresa Thompson, Vice President, Student Affairs and Enrollment Management Project Abstract, progression, and graduation. Methodology: The following items and their sources are included in this report

Hutcheon, James M.

49

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

50

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

51

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...

52

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.

53

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.

54

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.

55

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.

56

Regulatory analysis technical evaluation handbook. Final report  

SciTech Connect (OSTI)

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

NONE

1997-01-01T23:59:59.000Z

57

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

58

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

59

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

60

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

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


61

analysis process final: Topics by E-print Network  

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

Strategic Research and Analysis (OSRA Hutcheon, James M. 4 ALTERNATIVE JET FUEL SCENARIO ANALYSIS Final Report Energy Storage, Conversion and Utilization Websites Summary: America...

62

DOE Hydrogen Transition Analysis Workshop: Final Attendees List...  

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

Attendees List DOE Hydrogen Transition Analysis Workshop: Final Attendees List Attendee list for the DOE Hydrogen Transition Analysis Workshop on January 26, 2006....

63

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

64

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

65

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

66

Seismic Safety Margins Research Program. Phase I, final report. Major structure response (Project IV). Volume 5  

SciTech Connect (OSTI)

Task of the Major Structure Response Project within the Seismic Safety Margins Research Program (SSMRP) was to develop detailed finite element models of the Zion Nuclear Power Plant's containment building and auxiliary-fuel-turbine (AFT) complex. The resulting models served as input to the seismic methodology analysis chain.

Benda, B. J.; Johnson, J. J.; Lo, T. Y.

1981-05-01T23:59:59.000Z

67

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

68

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

69

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

SciTech Connect (OSTI)

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

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

2007-08-13T23:59:59.000Z

70

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

SciTech Connect (OSTI)

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

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

2011-12-15T23:59:59.000Z

71

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

72

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

73

ITER MHD stability analysis. Final report  

SciTech Connect (OSTI)

This report summarizes the final results and conclusions from work done for ITER under the DOE Task 18 (Raytheon Task ITER-GA 4002E). The work was performed in collaboration with D. Pearlstein and R. Bulmer of the Lawrence Livermore National Laboratory (LLNL), in close conjunction with D. Boucher of the ITER Joint Central Team (JCT). The work was partly done at General Atomics in San Diego and partly at LLNL. Approximately eight hours per week were spent from August 1994 through June 1995, with a no-cost extension through December 1995. The report covers work on the ideal MHD stability analysis for the ITER TAC scenarios and DIII-D ITER Demonstration Discharges, code modifications performed in order to efficiently and accurately complete the stability calculations, and additional collaborative efforts involving code benchmarking and dissemination of the DIII-D ITER Demonstration Discharge data. The work spawned several presentations and reports, including significant contributions to published IAEA Proceedings, and these are also summarized. 8 refs., 9 figs.

Turnbull, A.D.

1996-05-01T23:59:59.000Z

74

analysis study final: Topics by E-print Network  

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

University of 198 Distribution System Analysis Tools for Studying High Penetration of PV Power Transmission, Distribution and Plants Websites Summary: Final Project Report Power...

75

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network [OSTI]

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS.0 NEPA REQUIREMENTS: ENVIRONMENTAL IMPACTS OF THE ALTERNATIVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.1 Environmental Impacts of the Alternatives

76

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

77

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

78

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

SciTech Connect (OSTI)

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

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

1991-11-01T23:59:59.000Z

79

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

80

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

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


81

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

82

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.

83

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

84

The conservation planning analysis model. Final report  

SciTech Connect (OSTI)

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

Not Available

1993-12-31T23:59:59.000Z

85

Sandis irradiator for dried sewage solids. Final safety analysis report  

SciTech Connect (OSTI)

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

Morris, M.

1980-07-01T23:59:59.000Z

86

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

SciTech Connect (OSTI)

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

Miller, G.L.

1997-01-20T23:59:59.000Z

87

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

88

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

89

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

90

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

91

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

92

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

93

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

94

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

95

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...

96

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

SciTech Connect (OSTI)

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

NONE

1995-12-01T23:59:59.000Z

97

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

SciTech Connect (OSTI)

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

NONE

1995-12-01T23:59:59.000Z

98

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

99

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

100

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

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


101

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.

102

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.

103

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.

104

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

105

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

106

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

107

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS  

E-Print Network [OSTI]

ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS Amendment and Need The purpose of the non-AFA crab sideboard limits was to prevent vessels with crab QS from paper of all GOA sideboards for non-American Fisheries Act (AFA) crab vessels. In April 2007

108

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

109

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

110

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

111

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

SciTech Connect (OSTI)

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

N /A

2005-06-30T23:59:59.000Z

112

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

SciTech Connect (OSTI)

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

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

1997-05-01T23:59:59.000Z

113

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

114

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

115

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

116

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

SciTech Connect (OSTI)

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

Hecht, H.; Hecht, M.; Graff, S.; Green, W.; Lin, D.; Koch, S.; Tai, A.; Wendelboe, D. [SoHaR, Inc., Beverly Hills, CA (United States)

1996-06-01T23:59:59.000Z

117

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

118

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

119

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

120

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

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


121

Safety analysis report for packaging (Onsite) 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

122

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

SciTech Connect (OSTI)

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

Slavich, Antoinette; Daust, James E.

1999-10-01T23:59:59.000Z

123

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

124

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

125

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

126

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

127

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

128

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

129

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

130

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

131

UMTRA Project Office Federal Employee Occupational Safety and Health Program Plan. Final draft  

SciTech Connect (OSTI)

This document establishes the Federal Employee Occupational Safety and Health (FEOSH) Program for the US Department of Energy (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project Office. This program will ensure compliance with the applicable requirements of DOE Order 3790.1 B and DOE Albuquerque Operations Office (AL) Order 3790.1A. FEOSH Program responsibilities delegated by the DOE-AL to the UMTRA Project Office by AL Order 3790.1A also are assigned. The UMTRA Project Office has developed the UMTRA Project Environmental, Safety, and Health (ES&H) Plan (DOE, 1992), which establishes the basic programmatic ES&H requirements for all participants on the UMTRA Project. The ES&H plan is designed primarily to cover remedial action activities at UMTRA sites, and defines the ES&H responsibilities of both the UMTRA Project Office and its contractors. The UMTRA FEOSH Program described herein is a subset of the overall UMTRA ES&H program and covers only the federal employees working on the UMTRA Project.

Young, B.H.

1994-02-01T23:59:59.000Z

132

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

133

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

134

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

135

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

136

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

137

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

SciTech Connect (OSTI)

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

Golay, M.W.; Castillo, M.

1981-09-30T23:59:59.000Z

138

FINAL  

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 ProposedUsing ZirconiaPolicy andExsolutionFES Committees of VisitorsASCRReal-time2 FINAL

139

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...

140

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

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


141

The Independent Technical Analysis Process Final Report 2006-2007.  

SciTech Connect (OSTI)

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

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

2007-03-01T23:59:59.000Z

142

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

143

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

SciTech Connect (OSTI)

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

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

1998-09-01T23:59:59.000Z

144

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

145

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

146

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

147

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

148

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

SciTech Connect (OSTI)

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

Nuzum, J.L.

1997-01-31T23:59:59.000Z

149

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

150

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

E-Print Network [OSTI]

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

Griggs, D. P.

1982-01-01T23:59:59.000Z

151

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

152

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

153

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

154

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

155

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

156

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

157

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

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

this problem have been proposed, a final answer Source: Rosenbaum, Ren - Institute for Data Analysis and Visualization, University of California, Davis Collection: Computer...

158

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

SciTech Connect (OSTI)

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

Roy, G.J.

1992-09-01T23:59:59.000Z

159

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

160

Cost-effectiveness analysis of effluent standards and limitations for the metal finishing industry. Final report  

SciTech Connect (OSTI)

The report summarizes the results of a cost-effectiveness analysis of the metal finishing industry. The analysis considers the cost-effectiveness of the final metal finishing regulations for direct and indirect dischargers.

Not Available

1983-06-01T23:59:59.000Z

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


161

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

162

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

163

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

164

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

165

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

166

Analysis Of The Tank 6F Final Characterization Samples-2012  

SciTech Connect (OSTI)

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

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

2012-09-27T23:59:59.000Z

167

ANALYSIS OF THE TANK 6F FINAL CHARACTERIZATION SAMPLES-2012  

SciTech Connect (OSTI)

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

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

2012-06-28T23:59:59.000Z

168

Analysis of the Tank 6F Final Characterization Samples-2012  

SciTech Connect (OSTI)

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

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

2013-01-31T23:59:59.000Z

169

ANALYSIS OF THE TANK 5F FINAL CHARATERIZATION SAMPLES-2011  

SciTech Connect (OSTI)

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

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

2012-01-20T23:59:59.000Z

170

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.

171

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

172

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

173

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

174

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

175

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

176

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

177

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.

178

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.

179

analysis final technical: Topics by E-print Network  

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

use of components in software, which brings 12;encapsulation to the table Texas at San Antonio, University of 6 Computational Analysis of Technical Systems Mathematics Websites...

180

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

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


181

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

182

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

183

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

184

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

185

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

E-Print Network [OSTI]

Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

186

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

E-Print Network [OSTI]

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

Nero, A.V.

2010-01-01T23:59:59.000Z

187

Analysis of batteries for use in photovoltaic systems. Final report  

SciTech Connect (OSTI)

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

Podder, A; Kapner, M

1981-02-01T23:59:59.000Z

188

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

189

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

190

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

191

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

SciTech Connect (OSTI)

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

Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S. [SoHaR, Inc., Beverly Hills, CA (United States)

1997-10-01T23:59:59.000Z

192

FINAL AGENDA DOE Hydrogen Delivery Analysis and High Pressure Tanks  

E-Print Network [OSTI]

, National Renewable Energy Laboratory 3:00 BREAK 3:15 H2A Delivery Scenario Model and Analyses Jerry:15 Ethanol Delivery Cost and Distributed Ethanol Reforming Matt Ringer, National Renewable Energy Laboratory and breakfast 8:30 Hydrogen Delivery Infrastructure Analysis, Options and Trade-offs, Transition and Long

193

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

194

Hybrid methods for cybersecurity analysis : LDRD final report.  

SciTech Connect (OSTI)

Early 2010 saw a signi cant change in adversarial techniques aimed at network intrusion: a shift from malware delivered via email attachments toward the use of hidden, embedded hyperlinks to initiate sequences of downloads and interactions with web sites and network servers containing malicious software. Enterprise security groups were well poised and experienced in defending the former attacks, but the new types of attacks were larger in number, more challenging to detect, dynamic in nature, and required the development of new technologies and analytic capabilities. The Hybrid LDRD project was aimed at delivering new capabilities in large-scale data modeling and analysis to enterprise security operators and analysts and understanding the challenges of detection and prevention of emerging cybersecurity threats. Leveraging previous LDRD research e orts and capabilities in large-scale relational data analysis, large-scale discrete data analysis and visualization, and streaming data analysis, new modeling and analysis capabilities were quickly brought to bear on the problems in email phishing and spear phishing attacks in the Sandia enterprise security operational groups at the onset of the Hybrid project. As part of this project, a software development and deployment framework was created within the security analyst work ow tool sets to facilitate the delivery and testing of new capabilities as they became available, and machine learning algorithms were developed to address the challenge of dynamic threats. Furthermore, researchers from the Hybrid project were embedded in the security analyst groups for almost a full year, engaged in daily operational activities and routines, creating an atmosphere of trust and collaboration between the researchers and security personnel. The Hybrid project has altered the way that research ideas can be incorporated into the production environments of Sandias enterprise security groups, reducing time to deployment from months and years to hours and days for the application of new modeling and analysis capabilities to emerging threats. The development and deployment framework has been generalized into the Hybrid Framework and incor- porated into several LDRD, WFO, and DOE/CSL projects and proposals. And most importantly, the Hybrid project has provided Sandia security analysts with new, scalable, extensible analytic capabilities that have resulted in alerts not detectable using their previous work ow tool sets.

Davis, Warren Leon,; Dunlavy, Daniel M.

2014-01-01T23:59:59.000Z

195

Market analysis methodology: a utility case study. Final report  

SciTech Connect (OSTI)

The case study described in this report was conducted as part of EPRI Project RP1634 - Analytic Methods Used Outside the Electric Utility Industry. The primary objectives of the project were to: (1) explore planning and analysis techniques in use outside the utility industry, (2) identify those techniques which show promise for addressing utility issues, and (3) test them in actual utility situations to understand their real value, and the issues associated with adapting them to utility use.

Diamond, M.

1985-02-01T23:59:59.000Z

196

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

197

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

198

Comparative analysis of peat gasification reactor configuration. Final report  

SciTech Connect (OSTI)

This report summarizes the comparative analysis of two generic gasifiers (fluidized bed and entrained bed) and two specific hydrogasifiers (IGT's Peat-gas and Rockwell International's hydrogasifier). The objective is to establish a basis for recommending a peat gasification reactor configuration for testing in a DOE peat gasification pilot plant project. The approach involved the following four steps: data base evaluation, regression analysis, a chemical engineering evaluation of upstream and downstream equipment requirement, and computer simulation. Mathematical models and computer programs were developed to simulate the entrained-bed and fluidized-bed reactors. Parametric analyses were made, using these computer programs, to test the sensitivity and effects of significant operating variables (temperature, pressure and feed gas composition, for instance) on the product gas composition in the IGT gasifier and Rockwell International gasifier. This review and analysis concentrates primarily upon the Peatgas process of the Institute of Gas Technology and the Rockwell International Hydrogasification Process. The two-stage Peatgas process appears to have advantages in higher thermal efficiency, smaller capital investment, and its application of existing commercial operations and proven design practices from other types of fluidized solids processing units. There are two problems concerning operability of the Peatgas process: (1) the design of a fluidized solids unit for quite low standpipe densities (approximately 11 to 15 pounds per cubic foot), which may become even lower upon operating upsets, and (2) the potential problem of ash sintering in the very hot combustion zone of the steam-oxygen gasifier. On a relative time scale of development, the Peatgas would seem to be much closer to possible early commercialization than is the Rockwell system.

Not Available

1981-07-01T23:59:59.000Z

199

K Basin Hazard Analysis  

SciTech Connect (OSTI)

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

PECH, S.H.

2000-08-23T23:59:59.000Z

200

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

SciTech Connect (OSTI)

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

Not Available

1988-06-01T23:59:59.000Z

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


201

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

202

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

203

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

204

Caucasus Seismic Information Network: Data and Analysis Final Report  

SciTech Connect (OSTI)

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

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

2007-02-22T23:59:59.000Z

205

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

SciTech Connect (OSTI)

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

Bruce Wright

2011-08-30T23:59:59.000Z

206

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

207

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

208

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

209

Analysis of leaded and unleaded gasoline pricing. Final report  

SciTech Connect (OSTI)

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

Not Available

1985-03-15T23:59:59.000Z

210

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

211

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

212

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

SciTech Connect (OSTI)

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

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

1994-02-01T23:59:59.000Z

213

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

214

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

215

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

SciTech Connect (OSTI)

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

Not Available

1994-07-01T23:59:59.000Z

216

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

SciTech Connect (OSTI)

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

Not Available

1994-07-01T23:59:59.000Z

217

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

218

Environmental Assessment/Regulatory Impact Review/Final Regulatory Flexibility Analysis  

E-Print Network [OSTI]

Environmental Assessment/Regulatory Impact Review/Final Regulatory Flexibility Analysis (EA . . . . . . . . . . . 11 2.0 NEPA REQUIREMENTS: ENVIRONMENTAL IMPACTS OF THE ALTERNATIVES . . . . . . . . . . . . . . . . . . . . . . 15 2.1 Environmental Impacts of the Alternatives . . . . . 15 2.2 Whale watching activity in Alaska

219

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

SciTech Connect (OSTI)

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

Not Available

1994-08-01T23:59:59.000Z

220

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

SciTech Connect (OSTI)

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

Not Available

1994-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "final safety analysis" from the National Library of EnergyBeta (NLEBeta).
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221

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

222

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

223

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

224

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

SciTech Connect (OSTI)

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

NONE

1997-12-01T23:59:59.000Z

225

TIBER II/ETR final design report: Volume 3, 5. 0 Radiation safety and environment; 6. 0 Physics and technology R and D needs  

SciTech Connect (OSTI)

This paper discusses the design of the TIBER II Tokamak. This particular volume discusses: safety and environmental requirements and design targets; accident analyses; personnel safety and maintenance exposure; effluent control; waste management and decommissioning; safety considerations in building design; and safety and environmental conclusions and recommendations. (LSP)

Lee, J.D. (ed.)

1987-09-01T23:59:59.000Z

226

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...

227

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

228

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

229

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

SciTech Connect (OSTI)

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

NONE

1984-06-01T23:59:59.000Z

230

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

231

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

232

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

233

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

234

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

235

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

236

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

SciTech Connect (OSTI)

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

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

2011-05-01T23:59:59.000Z

237

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

238

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

239

Regulatory impact analysis of environmental standards for uranium mill tailings at active sites. Final report  

SciTech Connect (OSTI)

The Environmental Protection Agency was directed by Congress, under PL 95-604, the Uranium Mill Tailings Radiation Control Act of 1978, to set standards of general application that provide protection from the hazards associated with uranium mill tailings. Title I of the Act pertains to tailings at inactive sites for which the Agency has developed standards as part of a separate rulemaking. Title II of the Act requires standards covering the processing and disposal of byproduct materials at mills which are currently licensed by the appropriate regulatory authorities. This Regulatory Impact Analysis (RIA) addresses the standards developed under Title II. There are two major parts of the standards for active mills: standards for control of releases from tailings during processing operations and prior to final disposal, and standards for protection of the public after the disposal of tailings. This report presents a detailed analysis of standards for disposal only, since the analysis required for the operations standards is very limited.

Not Available

1983-03-01T23:59:59.000Z

240

Regulatory impact analysis of final environmental standards for uranium mill tailings at active sites  

SciTech Connect (OSTI)

The Environmental Protection Agency was directed by Congress, under PL 95-604, the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA), to set standards of general application that provide protection from the hazards associated with uranium mill tailings. Title II of the Act requires standards covering the processing and disposal of byproduct materials at mills which are currently licensed by the appropriate regulatory authorities. This Regulatory Impact Analysis (RIA) addresses the standards promulgated under Title II. There are two major parts of the standards for active mills: standards for control of releases from tailings during processing operations and prior to final disposal, and standards for protection of the public health and environment after the disposal of tailings. This report presents a detailed analysis of standards for disposal only, since the analysis required for the standards during mill operations is very limited.

Not Available

1983-09-01T23:59:59.000Z

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


241

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

242

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

243

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

SciTech Connect (OSTI)

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

Swain, A D; Guttmann, H E

1983-08-01T23:59:59.000Z

244

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

245

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

SciTech Connect (OSTI)

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

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

2009-03-14T23:59:59.000Z

246

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

Broader source: Energy.gov [DOE]

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

247

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

248

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

249

Activation Analysis of the Final Optics Assemblies at the National Ignition Facility  

SciTech Connect (OSTI)

Commissioning shots have commenced at the National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory. Within a year, the 192 laser beam facility will be operational and the experimental phase will begin. At each shot, the emitted neutrons will interact in the facility's surroundings, activating them, especially inside the target bay where the neutron flux is the highest. We are calculating the dose from those activated structures and objects in order to plan and minimize worker exposures during maintenance and normal NIF operation. This study presents the results of the activation analysis of the optics of the Final Optics Assemblies (FOA), which are a key contributor to worker exposure. Indeed, there are 48 FOAs weighting three tons each, and routine change-out and maintenance of optics and optics modules is expected. The neutron field has been characterized using the three-dimensional Monte Carlo particle transport code MCNP with subsequent activation analysis performed using the activation code, ALARA.

Dauffy, L S; Khater, H Y; Sitaraman, S; Brereton, S J

2008-10-14T23:59:59.000Z

250

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

251

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

252

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

253

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...

254

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...

255

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

SciTech Connect (OSTI)

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

Singh, M.; Energy Systems; TA Engineering

2008-02-29T23:59:59.000Z

256

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.

257

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

SciTech Connect (OSTI)

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

Eide, Steven Arvid; Thomas Wierman

2003-12-01T23:59:59.000Z

258

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

259

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

260

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

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


261

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

262

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

263

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

264

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

265

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

266

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

SciTech Connect (OSTI)

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

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

2010-04-30T23:59:59.000Z

267

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

268

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

SciTech Connect (OSTI)

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

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

2011-10-31T23:59:59.000Z

269

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

270

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

271

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.

272

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

SciTech Connect (OSTI)

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

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

2006-11-01T23:59:59.000Z

273

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...

274

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

275

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

276

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

E-Print Network [OSTI]

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

Chernowski, John

2009-01-01T23:59:59.000Z

277

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

E-Print Network [OSTI]

and calculation of population exposures (resulting These were the largest reactorsreactor, cannot be since, the definitive, particularly ably contain many of the calculationsreactor safety study, WASH-1400 of General background and objective Summary of methodology and results WASH-1400 Accident sequence identification and calculation

Nero, A.V.

2010-01-01T23:59:59.000Z

278

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

279

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

280

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

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281

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

282

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

283

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

284

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...

285

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

286

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

287

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

288

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

289

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

290

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

291

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

292

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

293

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

294

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

SciTech Connect (OSTI)

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

Vixie, Kevin R

2014-11-27T23:59:59.000Z

295

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

296

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

297

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

298

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

SciTech Connect (OSTI)

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

United States. Bonneville Power Administration.

1992-01-01T23:59:59.000Z

299

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

300

Three dimensional effects in analysis of PWR steam line break accident  

E-Print Network [OSTI]

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

Tsai, Chon-Kwo

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


301

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

302

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

303

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

304

Uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. Final report, February 16, 1990--December 31, 1994  

SciTech Connect (OSTI)

Dr. Robert Busch of the Department of Chemical and Nuclear Engineering was the principal investigator on this project with technical direction provided by the staff in the Nuclear Criticality Safety Group at Los Alamos. During the period of the contract, he had a number of graduate and undergraduate students working on subtasks. The objective of this work was to develop information on uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. During the first year of this project, most of the work was focused on setting up the SUN SPARC-1 Workstation and acquiring the literature which described the critical experiments. By august 1990, the Workstation was operational with the current version of TWODANT loaded on the system. MCNP, version 4 tape was made available from Los Alamos late in 1990. Various documents were acquired which provide the initial descriptions of the critical experiments under consideration as benchmarks. The next four years were spent working on various benchmark projects. A number of publications and presentations were made on this material. These are briefly discussed in this report.

Busch, R.D. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering

1995-02-24T23:59:59.000Z

305

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

306

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

307

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

308

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

309

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

310

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

311

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

312

Phase space analysis for three and four massive particles in final states  

E-Print Network [OSTI]

We propose formulae for computing the phase space integrals of $1\\to 3$ and $1\\to 4$ processes with massive particles in final states. As an application of these formulae we study the final state mass effects in some interesting phenomenological cases, giving fully integrated analytic results for the corresponding phase spaces. We consider also the $B_s-\\bar{B}_s$ process at NNLO and calculate one of the most complicated master integrals, which contributes to the $\\Delta\\Gamma_{B_s}$ at $O(\\alpha_s^2)$.

H. M. Asatrian; A. Hovhannisyan; A. Yeghiazaryan

2012-12-21T23:59:59.000Z

313

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

314

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

315

Procedure for conducting a human-reliability analysis for nuclear power plants. Final report  

SciTech Connect (OSTI)

This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study.

Bell, B.J.; Swain, A.D.

1983-05-01T23:59:59.000Z

316

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

317

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

318

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

319

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

320

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

SciTech Connect (OSTI)

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

Not Available

1994-01-01T23:59:59.000Z

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


321

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

322

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

323

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

324

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

SciTech Connect (OSTI)

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

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

1988-07-01T23:59:59.000Z

325

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

SciTech Connect (OSTI)

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

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

1988-07-01T23:59:59.000Z

326

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

SciTech Connect (OSTI)

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

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

1993-08-01T23:59:59.000Z

327

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

328

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.

329

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.

330

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.

331

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.

332

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.

333

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.

334

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.

335

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

SciTech Connect (OSTI)

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

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

2014-04-01T23:59:59.000Z

336

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

337

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

338

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

339

Affordability analysis of lead emission controls for a smelter-refinery. Final report  

SciTech Connect (OSTI)

This document evaluates the affordability and economic impact of additional control measures deemed necessary for a smelter-refinery to meet the lead emission standard. The emphasis in the analysis is on the impact of control costs on the smelter-refinery's profitability. The analysis was performed using control-cost data from two different lead-smelter studies in conjunction with other existing industry data.

Scherer, T.M.

1989-10-01T23:59:59.000Z

340

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

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


341

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

SciTech Connect (OSTI)

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

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

342

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

343

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

344

Site-specific earthquake response analysis for Paducah Gaseous Diffusion Plant, Paducah, Kentucky. Final report  

SciTech Connect (OSTI)

The Paducah Gaseous Diffusion Plant (PGDP), owned by the US Department of Energy (DOE) and operated under contract by Martin Marietta Energy systems, Inc., is located southwest of Paducah, Kentucky. An aerial photograph and an oblique sketch of the plant are shown in Figures 1 and 2, respectively. The fenced portion of the plant consists of 748 acres. This plant was constructed in the 1950`s and is one of only two gaseous diffusion plants in operation in the United States; the other is located near Portsmouth, Ohio. The facilities at PGDP are currently being evaluated for safety in response to natural seismic hazards. Design and evaluation guidelines to evaluate the effects of earthquakes and other natural hazards on DOE facilities follow probabilistic hazard models that have been outlined by Kennedy et al. (1990). Criteria also established by Kennedy et al. (1990) classify diffusion plants as ``moderate hazard`` facilities. The US Army Engineer Waterways Experiment Station (WES) was tasked to calculate the site response using site-specific design earthquake records developed by others and the results of previous geotechnical investigations. In all, six earthquake records at three hazard levels and four individual and one average soil columns were used.

Sykora, D.W.; Davis, J.J.

1993-08-01T23:59:59.000Z

345

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

SciTech Connect (OSTI)

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

Not Available

1982-09-01T23:59:59.000Z

346

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

347

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

348

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

349

Development of On-Board Fluid Analysis for the Mining Industry - Final report  

SciTech Connect (OSTI)

Pacific Northwest National Laboratory (PNNL: Operated by Battelle Memorial Institute for the Department of Energy) is working with the Department of Energy (DOE) to develop technology for the US mining industry. PNNL was awarded a three-year program to develop automated on-board/in-line or on-site oil analysis for the mining industry.

Pardini, Allan F.

2005-08-16T23:59:59.000Z

350

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

SciTech Connect (OSTI)

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

Elliott, D.C.

1986-06-01T23:59:59.000Z

351

Freedom car and vehicle technologies heavy vehicle program : FY 2007 benefits analysis, methodology and results -- final report.  

SciTech Connect (OSTI)

This report describes the approach to estimating the benefits and analysis results for the Heavy Vehicle Technologies activities of the FreedomCar and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identifying technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in subsequent activities. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. The market penetrations are used as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY07 Budget Request. The energy savings models are utilized by the FCVT program for internal project management purposes.

SIngh, M.; Energy Systems; TA Engineering

2008-02-29T23:59:59.000Z

352

FreedomCAR and vehicle technologies heavy vehicle program FY 2006. Benefits analysis : methodology and results - final report.  

SciTech Connect (OSTI)

This report describes the approach to estimating benefits and the analysis results for the Heavy Vehicle Technologies activities of the Freedom Car and Vehicle Technologies (FCVT) Program of EERE. The scope of the effort includes: (1) Characterizing baseline and advanced technology vehicles for Class 3-6 and Class 7 and 8 trucks, (2) Identification of technology goals associated with the DOE EERE programs, (3) Estimating the market potential of technologies that improve fuel efficiency and/or use alternative fuels, (4) Determining the petroleum and greenhouse gas emissions reductions associated with the advanced technologies. In FY 05 the Heavy Vehicles program activity expanded its technical involvement to more broadly address various sources of energy loss as compared to focusing more narrowly on engine efficiency and alternative fuels. This broadening of focus has continued in the activities planned for FY 06. These changes are the result of a planning effort that occurred during FY 04 and 05. (Ref. 1) This narrative describes characteristics of the heavy truck market as they relate to the analysis, a description of the analysis methodology (including a discussion of the models used to estimate market potential and benefits), and a presentation of the benefits estimated as a result of the adoption of the advanced technologies. These benefits estimates, along with market penetrations and other results, are then modeled as part of the EERE-wide integrated analysis to provide final benefit estimates reported in the FY06 Budget Request.

Singh, M.; Energy Systems; TA Engineering, Inc.

2006-01-31T23:59:59.000Z

353

Seismic Safety Margins Research Program: a concluding look  

SciTech Connect (OSTI)

The Seismic Safety Margins Research Program (SSMRP) was started in 1978 with the goal of developing tools and data bases to compute the probability of earthquake - caused radioactive release from commercial nuclear power plants. These tools and data bases were to help NRC to assess seismic safety at nuclear plants. The methodology to be used was finalized in 1982 and applied to the Zion Nuclear Power Station. The SSMRP will be completed this year with the development of a more simplified method of analysis and a demonstration of its use on Zion. This simplified method is also being applied to a boiling-water-reactor, LaSalle.

Cummings, G.E.

1984-01-01T23:59:59.000Z

354

Seismic risk analysis for General Electric Plutonium Facility, Pleasanton, California. Final report, part II  

SciTech Connect (OSTI)

This report is the second of a two part study addressing the seismic risk or hazard of the special nuclear materials (SNM) facility of the General Electric Vallecitos Nuclear Center at Pleasanton, California. The Part I companion to this report, dated July 31, 1978, presented the seismic hazard at the site that resulted from exposure to earthquakes on the Calaveras, Hayward, San Andreas and, additionally, from smaller unassociated earthquakes that could not be attributed to these specific faults. However, while this study was in progress, certain additional geologic information became available that could be interpreted in terms of the existance of a nearby fault. Although substantial geologic investigations were subsequently deployed, the existance of this postulated fault, called the Verona Fault, remained very controversial. The purpose of the Part II study was to assume the existance of such a capable fault and, under this assumption, to examine the loads that the fault could impose on the SNM facility. This report first reviews the geologic setting with a focus on specifying sufficient geologic parameters to characterize the postulated fault. The report next presents the methodology used to calculate the vibratory ground motion hazard. Because of the complexity of the fault geometry, a slightly different methodology is used here compared to the Part I report. This section ends with the results of the calculation applied to the SNM facility. Finally, the report presents the methodology and results of the rupture hazard calculation.

Not Available

1980-06-27T23:59:59.000Z

355

Final Harvest of Above-Ground Biomass and Allometric Analysis of the Aspen FACE Experiment  

SciTech Connect (OSTI)

The Aspen FACE experiment, located at the US Forest Service Harshaw Research Facility in Oneida County, Wisconsin, exposes the intact canopies of model trembling aspen forests to increased concentrations of atmospheric CO2 and O3. The first full year of treatments was 1998 and final year of elevated CO2 and O3 treatments is scheduled for 2009. This proposal is to conduct an intensive, analytical harvest of the above-ground parts of 24 trees from each of the 12, 30 m diameter treatment plots (total of 288 trees) during June, July & August 2009. This above-ground harvest will be carefully coordinated with the below-ground harvest proposed by D.F. Karnosky et al. (2008 proposal to DOE). We propose to dissect harvested trees according to annual height growth increment and organ (main stem, branch orders, and leaves) for calculation of above-ground biomass production and allometric comparisons among aspen clones, species, and treatments. Additionally, we will collect fine root samples for DNA fingerprinting to quantify biomass production of individual aspen clones. This work will produce a thorough characterization of above-ground tree and stand growth and allocation above ground, and, in conjunction with the below ground harvest, total tree and stand biomass production, allocation, and allometry.

Mark E. Kubiske

2013-04-15T23:59:59.000Z

356

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

357

An economic analysis of a monitored retrievable storage site for Tennessee. Final report and appendices  

SciTech Connect (OSTI)

The United States Department of Energy is charged with the task of identifying potential sites for a Monitored Retrievable Storage (MRS) Facility and reporting the results of its analysis to Congress by January 1986. DOE chose three finalist sites from 11 sites DOE analysts evaluated earlier. All three are in Tennessee, including two in Oak Ridge and one in Trousdale/Smith Counties. This paper is a summary of research undertaken on the economic effects of establishing the MRS facility in Tennessee. All three locations were considered in the analysis, but on some occasions attention is focused on the site preferred by DOE. The research was undertaken by the Center for Business and Economic Research (CBER), College of Business Administration, the University of Tennessee, Knoxville, under contract with the Tennessee Department of Economic and Community Development.

Fox, W.F.; Mayo, J.W.; Hansen, L.T.; Quindry, K.E.

1985-12-17T23:59:59.000Z

358

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

SciTech Connect (OSTI)

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

Lasseter, R.H.

1989-06-01T23:59:59.000Z

359

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:...

360

Recovery Planning for Endangered Salmon : A Multiple Attribute Analysis, Final Report,  

SciTech Connect (OSTI)

This analysis addresses multiple dimensions or attributes of recovery planning for endangered Snake River chinook stocks. The authors present a range of biological, economic, and social attributes for a number of recovery actions, and discuss aspects of the recovery actions that relate to each attribute. The emphasis on multiple attributes rather than on narrower biological measures alone reflects their belief that biological issues are only one of several sets of concerns that warrant attention in developing a recovery plan. Furthermore, the authors focus on both qualitative and quantitative factors because a lack of numerical information on certain attributes and recovery actions does not justify ignoring the non-numerical attributes or actions. After introducing the approach and providing the background, they define the attributes. An overview is provided of the biological modeling embedded in the analysis. The model used, the Stochastic Life Cycle Model (SLCM), determines the survival changes (relative to a base case) necessary to meet several biological criteria. These criteria reflect both the likelihood of population extinction and the projected population abundance 100 years into the future, relative to the initial abundance. The recovery options are outlined. The passage and harvest actions are characterized across the attributes is provided. The report assesses the life-stage survival improvements deemed necessary to avoid extinction and comments on the likelihood of meeting these improvements with the proposed actions. A cost-effectiveness analysis of the recovery strategies is provided.

Paulsen, Charles M.; Hyman, Jeffrey B.; Wernstedt, Kris

1993-12-09T23:59:59.000Z

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


361

Analysis of the need for intermediate and peaking technologies in the year 2000. Final report  

SciTech Connect (OSTI)

This analysis was conducted to assess the impact of load management on the future need for intermediate- and peak-generating technologies (IPTs) such as combustion turbines, pumped storage, and cycling coal plants. There would be a reduced need for IPTs if load-management activities such as time-of-use pricing, together with customer-owned energy-storage devices, hot-water-heater controls, and interruptible service can economically remove most of the variation from electric power demands. The objective of this analysis is to assess the need for IPTs in an uncertain future, which will probably include load management and time-differentiated electricity prices. The analysis is exploratory in nature and broad in scope. It does not attempt to predict the future or to model precisely the technical characteristics or economic desirability of load management. Rather, its purpose is to provide research and development planners with some basic insights into the order of magnitude of possible hourly demand shifts on a regional basis and to determine the impact of load management on daily and seasonal variations in electricity demand.

Barrager, S.M.; Campbell, G.L.

1980-04-01T23:59:59.000Z

362

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.

363

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

364

Reduced order models for thermal analysis : final report : LDRD Project No. 137807.  

SciTech Connect (OSTI)

This LDRD Senior's Council Project is focused on the development, implementation and evaluation of Reduced Order Models (ROM) for application in the thermal analysis of complex engineering problems. Two basic approaches to developing a ROM for combined thermal conduction and enclosure radiation problems are considered. As a prerequisite to a ROM a fully coupled solution method for conduction/radiation models is required; a parallel implementation is explored for this class of problems. High-fidelity models of large, complex systems are now used routinely to verify design and performance. However, there are applications where the high-fidelity model is too large to be used repetitively in a design mode. One such application is the design of a control system that oversees the functioning of the complex, high-fidelity model. Examples include control systems for manufacturing processes such as brazing and annealing furnaces as well as control systems for the thermal management of optical systems. A reduced order model (ROM) seeks to reduce the number of degrees of freedom needed to represent the overall behavior of the large system without a significant loss in accuracy. The reduction in the number of degrees of freedom of the ROM leads to immediate increases in computational efficiency and allows many design parameters and perturbations to be quickly and effectively evaluated. Reduced order models are routinely used in solid mechanics where techniques such as modal analysis have reached a high state of refinement. Similar techniques have recently been applied in standard thermal conduction problems e.g. though the general use of ROM for heat transfer is not yet widespread. One major difficulty with the development of ROM for general thermal analysis is the need to include the very nonlinear effects of enclosure radiation in many applications. Many ROM methods have considered only linear or mildly nonlinear problems. In the present study a reduced order model is considered for application to the combined problem of thermal conduction and enclosure radiation. The main objective is to develop a procedure that can be implemented in an existing thermal analysis code. The main analysis objective is to allow thermal controller software to be used in the design of a control system for a large optical system that resides with a complex radiation dominated enclosure. In the remainder of this section a brief outline of ROM methods is provided. The following chapter describes the fully coupled conduction/radiation method that is required prior to considering a ROM approach. Considerable effort was expended to implement and test the combined solution method; the ROM project ended shortly after the completion of this milestone and thus the ROM results are incomplete. The report concludes with some observations and recommendations.

Hogan, Roy E., Jr.; Gartling, David K.

2010-09-01T23:59:59.000Z

365

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

366

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

367

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

368

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

369

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

370

The European Large Area ISO Survey VIII: 90-micron final analysis and source counts  

E-Print Network [OSTI]

We present a re--analysis of the European Large Area ISO Survey (ELAIS) 90mum observations carried out with ISOPHOT, an instrument on board the ESA's Infrared Space Observatory (ISO). With more than 12 sq. deg., the ELAIS survey is the largest area covered by ISO in a single program and is about one order of magnitude deeper than the IRAS 100mum survey. The data analysis is presented and was mainly performed with the Phot Interactive Analysis software but using the pairwise method of Stickel et al. (2003) for signal processing from ERD (Edited Raw Data) to SCP (Signal per Chopper Plateau). The ELAIS 90mum catalogue contains 229 reliable sources with fluxes larger than 70 mJy and is available at www.blackwell-synergy.com. Number counts are presented and show an excess above the no-evolution model prediction. This confirms the strong evolution detected at shorter(15mum) and longer (170mum) wavelengths in other ISO surveys. The ELAIS counts are in agreement with previous works at 90mum and in particular with the deeper counts extracted from the Lockman hole observations. Comparison with recent evolutionary models show that the models of Franceschini et al. and Guiderdoni et al. which includes a heavily-extinguished population of galaxies give the best fit to the data. Deeper observations are nevertheless required to better discriminate between the model predictions in the far-infrared and are scheduled with the Spitzer Space Telescope which already started operating and will also be performed by ASTRO-F.

Ph. Heraudeau; S. Oliver; C. del Burgo; C. Kiss; M. Stickel; T. Mueller; M. Rowan-Robinson; A. Efstathiou; C. Surace; L. V. Toth; S. Serjeant; D. M. Alexander; A. Franceschini; D. Lemke; I. Perez-Fournon; T. Morel; J-L. Puget; D. Rigopoulou; B. Rocca-Volmerange; A. Verma

2004-10-15T23:59:59.000Z

371

Uncertainty analysis of the Measured Performance Rating (MPR) method. Final report  

SciTech Connect (OSTI)

A report was commissioned by the New York State Energy Research and Development Authority and the Electric Power Research Institute to evaluate the uncertainties in the energy monitoring method known as measured performance rating (MPR). The work is intended to help further development of the MPR system by quantitatively analyzing the uncertainties in estimates of the heat loss coefficients and heating system efficiencies. The analysis indicates that the MPR should detect as little as a 7 percent change in the heat loss coefficients and heating system efficiencies. The analysis indicate that the MPR should be able to detect as little as a 7 percent change in the heat loss coefficient at 95 percent confidence level. MPR appears sufficiently robust for characterizing common weatherization treatments; e.g., increasing attic insulation from R-7 to R-19 in a typical single-story, 1,100 sq. ft. house resulting in a 19 percent reduction in heat loss coefficient. Furnace efficiency uncertainties ranged up to three times those of the heat loss coefficients. Measurement uncertainties (at the 95 percent confidence level) were estimated to be from 1 to 5 percent for heat loss coefficients and 1.5 percent for a typical furnace efficiency. The analysis also shows a limitation in applying MPR to houses with heating ducts in slabs on grade and to those with very large thermal mass. Most of the uncertainties encountered in the study were due more to the methods of estimating the ``true`` heat loss coefficients, furnace efficiency, and furnace fuel consumption (by collecting fuel bills and simulating two actual houses) than to the MPR approach. These uncertainties in the true parameter values become evidence for arguments in favor of the need of empirical measures of heat loss coefficient and furnace efficiency, like the MPR method, rather than arguments against.

Not Available

1993-11-01T23:59:59.000Z

372

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

SciTech Connect (OSTI)

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

Nakhleh, Luay

2014-03-12T23:59:59.000Z

373

Partial-Wave Analysis of Centrally Produced Two-Pseudoscalar Final States in pp Reactions at COMPASS  

E-Print Network [OSTI]

COMPASS is a fixed-target experiment at the CERN SPS which focused on light-quark hadron spectroscopy during the data taking periods in 2008 and 2009. A world-leading data set was collected with a 190GeV/c hadron beam impinging on a liquid hydrogen target in order to study, inter alia, the central exclusive production of glueball candidates in the light-meson sector. Especially the double-Pomeron exchange mechanism is well suited for the production of mesons without valence quark content. We select centrally produced systems with two pseudo-scalar mesons in the final state from the COMPASS data set recorded with an incoming proton. The decay of this system is decomposed in terms of partial waves, where particular attention is paid to the inherent mathematical ambiguities of the amplitude analysis. Furthermore, we show that simple parametrisations are able to describe the mass dependence of the fit results with sensible Breit-Wigner parameters.

A. Austregesilo; for the COMPASS collaboration

2014-02-10T23:59:59.000Z

374

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

SciTech Connect (OSTI)

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

Potter, F.J.

1996-09-01T23:59:59.000Z

375

Analysis of lead-acid battery accelerated testing data. Final report  

SciTech Connect (OSTI)

Battelle conducted an independent review and analysis of the accelerated test procedures and test data obtained by Exide in the 3-year Phase I program to develop advanced lead-acid batteries for utility load leveling. Of special importance is the extensive data obtained in deep-discharge cycling tests on 60 cells at elevated temperatures over a 2-1/2 year period. The principal uncertainty in estimating cell life relates to projecting cycle life data at elevated temperature to the lower operating temperatures. The accelerated positive-grid corrosion test involving continuous overcharge at 50/sup 0/C provided some indication of the degree of grid corrosion that might be tolerable before failure. The accelerated positive-material shedding test was not examined in any detail. Recommendations are made for additional studies.

Clifford, J.E.; Thomas, R.E.

1983-08-01T23:59:59.000Z

376

Solar/gas systems impact analysis study. Final report, September 1982-July 1984  

SciTech Connect (OSTI)

The impacts of solar/gas technologies on gas consumers and on gas utilities was measured separately and compared against the impacts of competing gas and electric systems in four climatic regions of the U.S. A methodology was developed for measuring the benefits or penalties of solar/gas systems on a combined basis for consumers and distribution companies. The authors analysis shows that the combined benefits associated with solar/gas systems are generally greatest when the systems are purchased by customers who would have otherwise chosen high-efficiency electric systems (were solar/gas systems not available in the market place). The role of gas utilities in encouraging consumer acceptance of solar/gas systems was also examined in a qualitative fashion. The authors then developed a decision framework for analyzing the type and level of utility involvement in solar/gas technologies.

Hahn, E.F.; Preble, B.; Neill, C.P.; Loose, J.C.; Poe, T.E.

1984-07-01T23:59:59.000Z

377

Economic analysis of wind-powered farmhouse and farm building heating systems. Final report  

SciTech Connect (OSTI)

The study evaluated the break-even values of wind energy for selected farmhouses and farm buildings focusing on the effects of thermal storage on the use of WECS production and value. Farmhouse structural models include three types derived from a national survey - an older, a more modern, and a passive solar structure. The eight farm building applications that were analyzed include: poultry-layers, poultry-brooding/layers, poultry-broilers, poultry-turkeys, swine-farrowing, swine-growing/finishing, dairy, and lambing. These farm buildings represent the spectrum of animal types, heating energy use, and major contributions to national agricultural economic values. All energy analyses were based on hour-by-hour computations which allowed for growth of animals, sensible and latent heat production, and ventilation requirements. Hourly or three-hourly weather data obtained from the National Climatic Center was used for the nine chosen analysis sites, located throughout the United States and corresponding to regional agricultural production centers.

Stafford, R.W.; Greeb, F.J.; Smith, M.F.; Des Chenes, C.; Weaver, N.L.

1981-01-01T23:59:59.000Z

378

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

379

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

380

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

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


381

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

SciTech Connect (OSTI)

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

Esch, R.A.

1997-11-20T23:59:59.000Z

382

Energy Engineering Analysis (EEA) program for Lone Star Army Ammunition Plant, Texas. Executive summary. Final report  

SciTech Connect (OSTI)

The objective of this Energy Engineering Analysis (EEA) for LSAAP is threefold: Develop a systematic plan of projects which will result in reducing energy consumption. Consider renewable energy sources with the objective of establishing an orderly procedure for reducing use of non-renewable energy sources. Determine the feasibility of Total Energy (TE), Selective Energy (SE), and Central Heating Plant (CHP) concepts using alternative fuels. In essence, an assessment of the entire energy picture at LSAAP was undertaken. This report is a summary of that effort. LSAAP was originally built during 1941 and 1942 as a shell loading plant for the Army. After World War II, the facility was deactivated until 1951 when it was reactivated as a Government Owned, Contractor Operated (GOCO) facility. Day and Zimmerman was selected as the operator in 1951 and has been the operating contractor ever since. Located just west of Texarkana, Texas, LSAAP encompasses an area of approximately 15,546 acres. The primary mission of LSAAP is to load, assemble and pack ammunition and ammunition components for the Army.

NONE

1981-12-31T23:59:59.000Z

383

High Btu gas from peat. A feasibility study. Part 3. Market analysis. Task 8. Final report  

SciTech Connect (OSTI)

The primary objective of this task, which was the responsibility of the Minnesota Gas Company, was to identify and characterize the market potential for the plant by-products - BTX (mixture of benzene, toluene and xylene), phenol, ammonia, sulfur, and sodium sulfate - and to assign value to them. Although traditionally a growth industry, the chemicals market has been generally weakened by the recession, and is experiencing back to back years of declining production. This is due to bad health of specific end uses, such as fertilizer from ammonia. In the long run, this trend is expected to moderate. It is felt that the proposed peat plant has a favorable position in the markets of each of its by-products. This is due to the synergism with nearby industries which are major consumers of these by-products. In the case of sulfur and ammonia, the Red River agricultural area is a large potential market. For sodium sulfate, phenols and perhaps BTX, the nearby paper and timber products industries are large potential markets. The values for these by-products used in the financial analysis were intentionally conservative. This is because of the uncertainty in the quantity and quality. More tests are needed in an integrated facility in order to determine these factors and the variability of each. This is particularly true of the by-product oils which could vary significantly with operating conditions and may even require alternate processing schemes. 18 references, 9 figures, 14 tables.

Not Available

1982-01-01T23:59:59.000Z

384

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

SciTech Connect (OSTI)

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

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

2012-05-31T23:59:59.000Z

385

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

SciTech Connect (OSTI)

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

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

1980-12-01T23:59:59.000Z

386

A regulatory analysis on emergency preparedness for fuel cycle and other radioactive material licensees: Final report  

SciTech Connect (OSTI)

The question this Regulatory Analysis sought to answer is: should the NRC impose additional emergency preparedness requirements on certain fuel cycle and other radioactive material licensees for dealing with accidents that might have offsite releases of radioactive material. To answer the question, we analyzed potential accidents for 15 types of fuel cycle and other radioactive material licensees. An appropriate plan would: (1) identify accidents for which protective actions should be taken by people offsite; (2) list the licensee's responsibilities for each type of accident, including notification of local authorities (fire and police generally); and (3) give sample messages for local authorities including protective action recommendations. This approach more closely follows the approach used for research reactors than for power reactors. The low potential offsite doses (acute fatalities and injuries not possible except possibly for UF/sub 6/ releases), the small areas where actions would be warranted, the small number of people involved, and the fact that the local police and fire departments would be doing essentially the same things they normally do, are all factors that tend to make a simple plan adequate. This report discusses the potentially hazardous accidents, and the likely effects of these accidents in terms of personnel danger.

McGuire, S.A.

1988-01-01T23:59:59.000Z

387

Final Report  

SciTech Connect (OSTI)

OAK-B135 This is the final report from the project Hydrodynamics by High-Energy-Density Plasma Flow and Hydrodynamics and Radiation Hydrodynamics with Astrophysical Applications. This project supported a group at the University of Michigan in the invention, design, performance, and analysis of experiments using high-energy-density research facilities. The experiments explored compressible nonlinear hydrodynamics, in particular at decelerating interfaces, and the radiation hydrodynamics of strong shock waves. It has application to supernovae, astrophysical jets, shock-cloud interactions, and radiative shock waves.

R Paul Drake

2004-01-12T23:59:59.000Z

388

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

389

Analysis of production line motor failure. CRADA final report for CRADA number Y-1293-0215  

SciTech Connect (OSTI)

The Oak Ridge National Laboratory (ORNL) was approached by a Food Products Manufacturer (FPM) to investigate the rapid failure of motors in a manufacturing facility. It was reported that some motors or their bearings were being replaced after as little as four months of service. The deciding symptom for replacement was always high motor vibration. To protect against unscheduled downtime in the middle of a process run, the FPM`s maintenance team removes a motor from service when its vibration level reaches a conservative threshold of approximately 0.4 inches per second. In their experience, motors left in service after reaching this vibration threshold can fail at any time within the time span of the next process run causing significant losses of raw material and production capacity. A peculiar finding of vibration level trend analysis was that at least one motor exhibited cyclic variations with 24-hour periodicity. The vibration level reached a maximum at about 4:00 a.m., ramped down during the day, and then rose again during the night. Another peculiarity was that most of the vibration energy in the affected motors was at the 120 Hz frequency. Since this is twice the 60 Hz line frequency the FPM suspected the vibration was electrically induced. The electric loads at the FPMs plant remain constant during the five days of a continuous production run. Thus, the periodicity of the vibration observed, with its daily peaking at about four am, suggested the possibility of being driven by changes in the electrical power grid external to the plant.

Kueck, J. [Lockheed Martin Energy Systems, Inc., Oak Ridge, TN (United States); Talbott, C. [M& M Mars, Inc., Chicago, IL (United States)

1995-02-10T23:59:59.000Z

390

Availability analysis of an integrated gasification-combined cycle: Final report  

SciTech Connect (OSTI)

The Electric Power Research Institute (EPRI) contracted with ARINC Research Corporation to perform availability assessments of an integrated coal gasification-combined-cycle (IGCC) design. The objective of the study was to quantify the availability impact associated with several design and operating options specified by EPRI. In addition, several scheduled maintenance options for the IGCC plant were evaluated. The IGCC plant addressed in this analysis employs many modular design features that give the plant high equivalent availability through redundancy. The study focused on evaluating and quantifying the expected changes in unit capability, equivalent availability, and heat rate associated with various design alternatives. The findings of the baseline case studies are as follows: (1) The Baseline IGCC design using four gasifiers with 11.2% spare gasification capacity and three combustion turbine/HRSGs sets will have an expected equivalent availability of 86.18% and an average heat rate of 9002 Btu/kWh. (2) The Baseline with Supplemental Firing design using four gasifiers with the 11.2% spare gasification capacity being used to produce supplemental steam and with three combustion turbine HRSG sets will have an expected equivalent availability of 85.64% and an average heat rate of 9147 Btu/kWh. (3) The Baseline with Natural Gas Backup design using four gasifiers and three combustion turbine/HRSG sets with supplemental natural gas backup will have an expected equivalent availability of 91.53% with an average heat rate of 8981 Btu/kWh and a coal-to-natural gas fuel mixture of 23:1. 49 figs., 66 tabs.

Not Available

1987-06-01T23:59:59.000Z

391

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...

392

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

393

Safety Values  

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

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

394

Radiation Safety  

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

Weeks of training * 15 of that is OJT * General Code of Operating Rules * Air Brake & Train Handling * System Special Instructions * Safety Instructions * Federal Regulations *...

395

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

E-Print Network [OSTI]

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

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

2005-08-05T23:59:59.000Z

396

Nuclear Safety Analysis Reports  

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

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

1992-04-30T23:59:59.000Z

397

antimalarial drug safety: Topics by E-print Network  

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

of fault tree analysis Reif, Wolfgang 404 Toolbox Safety Talk Fume Hood Decommissioning Biology and Medicine Websites Summary: Toolbox Safety Talk Fume Hood...

398

anesthesia patient safety: Topics by E-print Network  

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

of fault tree analysis Reif, Wolfgang 451 Toolbox Safety Talk Fume Hood Decommissioning Biology and Medicine Websites Summary: Toolbox Safety Talk Fume Hood...

399

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

SciTech Connect (OSTI)

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

West, M H

2008-12-17T23:59:59.000Z

400

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

E-Print Network [OSTI]

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

Lee, John D.

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


401

Precursor systems analyses of automated highway systems: Commercial vehicle and transit ahs analysis. Volume 6. Final report, September 1993-February 1995  

SciTech Connect (OSTI)

This document is the final report of the Automated Highway System (AHS). The activities of Commercial Vehicle and Transit AHS Analysis are reported on in this document. This document type is resource materials. This volume is the six in a series. There are nine other volumes in the series.

Bottiger, F.; Chemnitz, H.D.; Doorman, J.; Franke, U.; Zimmerman, T.

1995-03-01T23:59:59.000Z

402

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

403

Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report  

SciTech Connect (OSTI)

For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial temperature distribution, heat exchanger conditions. The facility also has potential to integrate additional components of the prototype system, such as intermediate thermal-hydraulics loops, real-time grid-demand data, energy storage, and HTSE.

Shannon Bragg-Sitton; Piyush Sabharwall; Rohit Upadhya

2014-01-01T23:59:59.000Z

404

Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report  

SciTech Connect (OSTI)

For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial temperature distribution, heat exchanger conditions. The facility also has potential to integrate additional components of the prototype system, such as intermediate thermal-hydraulics loops, real-time grid-demand data, energy storage, and HTSE.

Shannon Bragg-Sitton; Piyush Sabharwall; Rohit Upadhya

2013-08-01T23:59:59.000Z

405

Idaho National Laboratory Safety Presentations  

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

* Hand Tool Safety * Protect Your Hearing * Water Safety * Home Firearms Safety * Bicycle Safety * Pedestrian Safety * Others Outdoor Survival Safety (K-Middle School) What to...

406

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

SciTech Connect (OSTI)

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

Rathbun, R.

1994-03-02T23:59:59.000Z

407

Facility Safety  

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

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

2002-05-20T23:59:59.000Z

408

Updated laser safety&hazard analysis for the ARES laser system based on the 2007 ANSI Z136.1 standard.  

SciTech Connect (OSTI)

A laser safety and hazard analysis was performed for the temperature stabilized Big Sky Laser Technology (BSLT) laser central to the ARES system based on the 2007 version of the American National Standards Institute's (ANSI) Standard Z136.1, for Safe Use of Lasers and the 2005 version of the ANSI Standard Z136.6, for Safe Use of Lasers Outdoors. The ARES laser system is a Van/Truck based mobile platform, which is used to perform laser interaction experiments and tests at various national test sites.

Augustoni, Arnold L.

2007-08-01T23:59:59.000Z

409

Toolbox Safety Talk Hot Work Safety Procedures  

E-Print Network [OSTI]

Toolbox Safety Talk Hot Work Safety Procedures Environmental Health & Safety Facilities Safety-in sheet to Environmental Health & Safety for recordkeeping. "Hot Work" is defined as any temporary WORK Obtain a hot work permit from your supervisor or safety rep. Ensure fire/smoke detection

Pawlowski, Wojtek

410

Toolbox Safety Talk Machine Shop Safety  

E-Print Network [OSTI]

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

Pawlowski, Wojtek

411

ENVIRONMENTAL HEALTH AND SAFETY GENERAL SAFETY MANUAL  

E-Print Network [OSTI]

ENVIRONMENTAL HEALTH AND SAFETY GENERAL SAFETY MANUAL May 10, 2002 #12;i Acknowledgements Environmental Health and Safety gratefully acknowledges the assistance provided by the University Safety Council extremely helpful. #12;ii Environmental Health and Safety General Safety Manual Table of Contents Section

Maroncelli, Mark

412

LASER SAFETY POLICY MANUAL ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network [OSTI]

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

Houston, Paul L.

413

Final technical report for Phenomic Analysis of Natural and Induced Variation in Brachypodium Distachyon DE-SC0001526  

SciTech Connect (OSTI)

The goal of this project was to apply high-throughput, non-destructive phenotyping (phenomics) to collections of natural variants and induced mutants of the model grass Brachypodium distachyon and characterize a small subset of that material in detail. B. distachyon is well suited to this phenomic approach because its small size and rapid generation time allow researchers to grow many plants under carefully controlled conditions. In addition, the simple diploid genetics, high quality genome sequence and existence of numerous experimental tools available for B. distachyon allow us to rapidly identify genes affecting specific phenotypes. Our phenomic analysis revealed great diversity in biofuel-relevant traits like growth rate, biomass and photosynthetic rate. This clearly demonstrated the feasibility of applying a phenomic approach to the model grass B. distachyon. We also demonstrated the utility of B. distachyon for studying mature root system, something that is virtually impossible to do with biomass crops. We showed tremendous natural variation in root architecture that can potentially be used to design crops with superior nutrient and water harvesting capability. Finally, we demonstrated the speed with which we can link specific genes to specific phenotypes by studying two mutants in detail. Importantly, in both cases, the specific biological lessons learned were grass-specific and could not have been learned from a dicot model system. Furthermore, one of the genes affects cell wall integrity and thus may be a useful target in the context of biomass crop improvement. Ultimately, all this information can be used to accelerate the creation of improved biomass crops.

Vogel, John P.

2014-12-17T23:59:59.000Z

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

DOE HANDBOOK ELECTRICAL SAFETY  

E-Print Network [OSTI]

DOE HANDBOOK ELECTRICAL SAFETY U.S. Department of Energy AREA SAFT Washington, D.C. 20585 of 139 3.0 HAZARD ANALYSIS 3.1 INTRODUCTION This chapter provides tools for assessing electrical hazards). The risk of a worker to an exposed electrical hazard is determined by (a) the classification

416

Safety, Security & Fire Report  

E-Print Network [OSTI]

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

Straight, Aaron

417

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors  

SciTech Connect (OSTI)

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Sekimoto, H., E-mail: hsekimot@gmail.com [Research Lab. For Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo (Japan)

2014-09-30T23:59:59.000Z

418

IMPORTANCE OF SAFETY CULTURE ASSESSMENT  

SciTech Connect (OSTI)

Safety Management has lately been considered by some Nuclear Regulatory agencies as the tool on which to concentrate their efforts to implement modern regulation structures, because Safety Culture was said to be difficult to monitor. However, Safety Culture can be assessed and monitored even if it is problematical to make Safety Culture the object of regulation. This paper stresses the feasibility and importance of Safety Culture Assessment based on self-assessment applications performed in several nuclear organizations in Latin America. Reasons and ownership for assessing Safety Culture are discussed, and relevant aspects considered for setting up and programming such an assessment are shown. Basic principles that were taken into account, as well as financial and human resources used in actual self-assessments are reviewed, including the importance of adequate statistical analyses and the necessity of proper feed-back of results. The setting up of action plans to enhance Safety Culture is the final step of the assessment program that once implemented will enable to establish a Safety Culture monitoring process within the organization.

Spitalnik, J.

2004-10-06T23:59:59.000Z

419

Spent nuclear fuel project cold vacuum drying facility safety equipment list  

SciTech Connect (OSTI)

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

IRWIN, J.J.

1999-02-24T23:59:59.000Z

420

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

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


421

Technical Review Report for the Model 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

422

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

423

ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network [OSTI]

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

Brownstone, Rob

424

Toolbox Safety Talk Welding & Metal Work Safety  

E-Print Network [OSTI]

Toolbox Safety Talk Welding & Metal Work Safety Environmental Health & Safety Facilities Safety or harmful emission giving metals. Welding Safety When welding outside of a designated welding booth, ensure injury. Avoid welding on materials such as galvanized or stainless steel in order to minimize toxic fume

Pawlowski, Wojtek

425

MINIMARS conceptual design: Final report  

SciTech Connect (OSTI)

This volume contains the following sections: (1) fueling systems; (2) blanket; (3) alternative blanket concepts; (4) halo scraper/direct converter system study and final conceptual design; (5) heat-transport and power-conversion systems; (6) tritium systems; (7) minimars air detritiation system; (8) appropriate radiological safety design criteria; and (9) cost estimate. (MOW)

Lee, J.D. (ed.)

1986-09-01T23:59:59.000Z

426

Gas Pipeline Safety (Indiana)  

Broader source: Energy.gov [DOE]

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

427

Electrical Safety Committee Charter  

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

ANL Electrical Safety Page DOE Electrical Safety Handbook General Statement Home & Office Equipment Statement APS Electrical Safety Update Guidelines for Working on Voltages < 240...

428

Safety Overview Committee (SOC)  

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

(SOC) Charter 1. Purpose The Safety Overview Committee establishes safety policies and ad hoc safety committees. 2. Membership Membership will include the following individuals:...

429

Nuclear Safety Regulatory Framework  

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

overall Nuclear Safety Policy & ESH Goals Safety Basis Review and Approval In the DOE governance model, contractors responsible for the facility develop the safety basis and...

430

Asymptotic Safety  

E-Print Network [OSTI]

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

R. Percacci

2008-11-18T23:59:59.000Z

431

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

SciTech Connect (OSTI)

Large-scale simulation datasets can be modeled as high-dimensional scalar functions defined over a discrete sample of the domain. The goals of our proposed research are two-fold. First, we would like to provide structural analysis of a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. TopoXG is a software package that is designed to address these goals. The unique contribution of TopoXG lies in exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to TopoXG, by highlighting its analysis and visualization capabilities, and giving several use cases involving datasets from nuclear reactor safety simulations.

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

2012-10-01T23:59:59.000Z

432

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,...

433

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

SciTech Connect (OSTI)

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

Chou, P

2011-12-14T23:59:59.000Z

434

Machine Shop Safety Tips & Safety Guidelines GENERAL SAFETY TIPS  

E-Print Network [OSTI]

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

Veiga, Pedro Manuel Barbosa

435

Safety basis academy summary of project implementation from 2007-2009  

SciTech Connect (OSTI)

During fiscal years 2007 through 2009, in accordance with Performance Based Incentives with DOE/NNSA Los Alamos Site Office, Los Alamos National Security (LANS) implemented and operated a Safety Basis Academy (SBA) to facilitate uniformity in technical qualifications of safety basis professionals across the nuclear weapons complex. The implementation phase of the Safety Basis Academy required development, delivery, and finalizing a set of 23 courses. The courses developed are capable of supporting qualification efforts for both federal and contractor personnel throughout the DOE/NNSA Complex. The LANS Associate Director for Nuclear and High Hazard Operations (AD-NHHO) delegated project responsibillity to the Safety Basis Division. The project was assigned to the Safety Basis Technical Services (SB-TS) Group at Los Alamos National Laboratory (LANL). The main tasks were project needs analysis, design, development, implementation of instructional delivery, and evaluation of SBA courses. DOE/NNSA responsibility for oversight of the SBA project was assigned to the Chief of Defense for Nuclear Safety, and delegated to the Authorization Basis Senior Advisor, Continuous Learning Chair (CDNS-ABSA/CLC). NNSA developed a memorandum of agreement with LANS AD-NHHO. Through a memorandum of agreement initiated by NNSA, the DOE National Training Center (NTC) will maintain the set of Safety Basis Academy courses and is able to facilitate course delivery throughout the DOE Complex.

Johnston, Julie A [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

436

OCCUPATIONAL SAFETY and HEALTH  

E-Print Network [OSTI]

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

Weaver, Harold A. "Hal"

437

OCCUPATIONAL HEALTH AND SAFETY  

E-Print Network [OSTI]

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

438

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

439

Safety harness  

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

440

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

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


441

Safety valve  

DOE Patents [OSTI]

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

Bergman, Ulf C. (Malmoe, SE)

1984-01-01T23:59:59.000Z

442

Russell Furr Laboratory Safety &  

E-Print Network [OSTI]

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

443

Studies of the analyte-carrier interface in flow injection analysis. Final report, June 1, 1987--January 31, 1992  

SciTech Connect (OSTI)

Chemical analysis in flowing solution is popular for automation of classical methods. However, most of the classical methods are not specific enough for direct multicomponent analysis of simple mixtures. This research project has the goals of study of rapid multicomponent analysis of transient species in flowing media, and investigations of chemical reactions at interfaces and of effects of competition on distribution of products from interfacial reaction. This report summarizes work done over the past 4.5 years; support has been terminated.

Brown, S.D.

1992-12-31T23:59:59.000Z

444

Proposed Tenaska Washington II Generation Project : Final Environmental Impact Statement. Volume 1: Environmental Analysis and Technical Appendices.  

SciTech Connect (OSTI)

BPA is considering whether to purchase electrical power from a proposed privately-owned combustion-turbine electrical generation plant in Washington. The plant would be fired by natural gas and would use combined-cycle technology to generate 240 average megawatts (aMW) of energy. The plant would be developed, owned, and operated by Tenaska Washington Partners II, L.P. The project would be located about 19 kilometers (12 miles) southeast of downtown Tacoma in the Frederickson Industrial Area, Pierce County. The proposed plant would occupy about half of a 6.4-hectare (16-acre) parcel and would be consistent with the industrial character of its surroundings. The proposed site is currently undeveloped and zoned for industrial use by the county. Main environmental concerns identified in the scoping process and in comments on the Draft Environmental Impact Statement (EIS) include: (1) potential air quality impacts, such as emissions and their contribution to the {open_quotes}greenhouse{close_quotes} effect; (2) potential health and safety impacts, such as nuisance odors, plant safety, visibility and heat-emission systems which may affect low-flying planes and potential health effects of electric and magnetic fields; and (3) potential water quality and quantity impacts, such as the amount of wastewater to be discharged, the source and amount of water required for plant operation. These and other issues are discussed in detail in the EIS. The proposed project already includes many features designed to reduce environmental impacts. Based on investigations performed for the EIS, no significant unavoidable adverse environmental impacts associated with the proposed project were identified, and no evidence emerged to suggest that the proposed action is controversial. The EIS is being mailed to numerous agencies, groups, and individuals (see Section 8.0). There will be a 30-day no-action period before any decisions are made and the Record of Decision is signed.

United States. Bonneville Power Administration.

1994-01-01T23:59:59.000Z

445

MCNP6 Results for the Phase III Sensitivity Benchmark of the OCED/NEA Expert Group on Uncertainty Analysis for Criticality Safety Assessment  

SciTech Connect (OSTI)

Within the last decade, there has been increasing interest in the calculation of cross section sensitivity coefficients of k{sub eff} for integral experiment design and uncertainty analysis. The OECD/NEA has an Expert Group devoted to Sensitivity and Uncertainty Analysis within the Working Party for Nuclear Criticality Safety. This expert group has developed benchmarks to assess code capabilities and performance for doing sensitivity and uncertainty analysis. Phase III of a set of sensitivity benchmarks evaluates capabilities for computing sensitivity coefficients. MCNP6 has the capability to compute cross section sensitivities for k{sub eff} using continuous-energy physics. To help verify this capability, results for the Phase III benchmark cases are generated and submitted to the Expert Group for comparison. The Phase III benchmark has three cases: III.1, an array of MOX fuel pins, III.2, a series of infinite lattices of MOX fuel pins with varying pitches, and III.3 two spheres with homogeneous mixtures of UF{sub 4} and polyethylene with different enrichments.

Kiedrowski, Brian C. [Los Alamos National Laboratory

2012-06-19T23:59:59.000Z

446