Sample records for accident conditions view

  1. accident conditions final: Topics by E-print Network

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

    to evaluate human weather discomfort due to hot conditions and then tested for work accident differences using non-parametric procedures. Present findings showed that hot weather...

  2. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect (OSTI)

    El-Genk, Mohamed

    2012-10-19T23:59:59.000Z

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air ?helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  3. accident conditions key: Topics by E-print Network

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

    accident conditions key First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Range Condition: Key to...

  4. accident ria conditions: Topics by E-print Network

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

    accident ria conditions First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Ria-kehitys Ruby Rsence...

  5. accident conditions vercors: Topics by E-print Network

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

    accident conditions vercors First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 RUNNING HEAD: Subaerial...

  6. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect (OSTI)

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01T23:59:59.000Z

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  7. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    SciTech Connect (OSTI)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01T23:59:59.000Z

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  8. Technical basis document for the release from contaminated facility representative accident and associated represented hazardous conditions

    SciTech Connect (OSTI)

    OBERG, B.D.

    2003-03-22T23:59:59.000Z

    This document supports the Tank Farms Documented Safety Analysis and describes the risk binning process and the technical basis for assigning risk bins for the release from contaminated facility representative accident and associated represented hazardous conditions. The representative accidents qualitatively considered are fires, deflagrations, and load drops in contaminated areas. The risks from a separate evaluation of compressed gas hazards are also summarized.

  9. Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages

    SciTech Connect (OSTI)

    MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

    2003-01-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

  10. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GRIGSBY, J.M.

    2005-03-03T23:59:59.000Z

    Technical Basis Document for the Nuclear Criticality Representative Accident and Associate Represented Hazardous Conditions. Revision 2 of RPP-12371 provides accident consequence estimates for a hypothetical criticality event in an above grade facility (e.g. DBVS, CH-TRUM, and S-109 PWRS). This technical basis document was developed to support RPP-13033, ''Tank Farms Documented Safety Analysis (DSA)'', and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  11. Iodine behavior in containment under LWR accident conditions

    SciTech Connect (OSTI)

    Wisbey, S.J.; Beahm, E.C.; Shockley, W.E.; Wang, Y.M.

    1986-01-01T23:59:59.000Z

    The description of containment iodine behavior in reactor accident sequences requires an understanding of iodine volatility effects, deposition and revaporization/resuspension (from surfaces and aerosols), chemical changes between species, and mass transport. The experimental work in this program has largely centered on the interactions of iodine in or with water pools. The formation of volatile iodine, as I/sub 2/ or organic iodides, is primarily dependent on radiation and solution pH. Lower pH results in increased formation of volatile iodine species; thus, for example, a pH of 3.05 resulted in a conversion of I/sup -/ to I/sub 2/ that was more than two orders of magnitude greater than tests run at pH 6.1 or 6.8. The formation or organic iodides involving water pools has been linked to the presence of iodine as I/sub 2/, the solution/gas contact, and to the type of organic material.

  12. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-06-17T23:59:59.000Z

    This document was developed to support the documented safety analysis (DSA) and describes the process and basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. Revision 1 incorporates ORP IRT comments to enhance the technical presentation and also makes editorial changes. This technical basis document was developed to support the documented safety analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence.

  13. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect (OSTI)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01T23:59:59.000Z

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  14. Technical basis for the aboveground structure failure accident & associated represented hazardous conditions

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-05-15T23:59:59.000Z

    This technical basis document describes the risk binning process and the technical basis for assigning risk bins for the above-ground structure failure representative accident and associated represented hazardous conditions. This document was developed to support the documented safety analysis.

  15. PRESSURE INTEGRITY OF 3013 CONTAINER UNDER POSTULATED ACCIDENT CONDITIONS

    SciTech Connect (OSTI)

    Rawls, G.

    2010-02-01T23:59:59.000Z

    A series of tests was carried out to determine the threshold for deflagration-to-detonation transition (DDT), structural loading, and structural response of the Department of Energy 3013 storage systems for the case of an accidental explosion of evolved gas within the storage containers. Three experimental fixtures were used to examine the various issues and three mixtures consisting of either stoichiometric hydrogen-oxygen, stoichiometric hydrogen-oxygen with added nitrogen, or stoichiometric hydrogen-oxygen with an added nitrogen-helium mixture were tested. Tests were carried out as a function of initial pressure from 1 to 3.5 bar and initial temperature from room temperature to 150 C. The elevated temperature tests resulted in a slight increase in the threshold pressure for DDT. The elevated temperature tests were performed to ensure the test results were bounding. Because the change was not significant the elevated temperature data are not presented in the paper. The explosions were initiated with either a small spark or a hot surface. Based on the results of these tests under the conditions investigated, it can be concluded that DDT of a stoichiometric hydrogen-oxygen mixture (and mixtures diluted with nitrogen and helium) within the 3013 containment system does not pose a threat to the structural integrity of the outer container.

  16. Technical basis for the nuclear criticality representative accident and associated represented hazardous conditions

    SciTech Connect (OSTI)

    CARSON, D.M.

    2003-03-20T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  17. Technical basis for the tank bump representative accident and associated hazardous conditions

    SciTech Connect (OSTI)

    WILLIAMS, J.C.

    2003-03-21T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for the tank bump representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', as described in this report.

  18. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    SciTech Connect (OSTI)

    Nourbakhsh, H.P. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01T23:59:59.000Z

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented.

  19. Improved assessment of population doses and risk factors for a nuclear power plant under accident conditions 

    E-Print Network [OSTI]

    Meyer, Christopher Martin

    1985-01-01T23:59:59.000Z

    of the requirements for the degree of MASTER OF SCIENCE August 1985 Major Subject: Nuclear Engineering IMPROVED ASSESSMENT OF POPULATION DOSES AND RISK FACTORS FOR A NUCLEAR POWER PLANT UNDER ACCIDENT CONDITIONS A Thesis by CHRISTOPHER MARTIN MEYER Approved... as to style and content by: G. A. Schlapper (Chair of Committee R. B. Ko zen (Member) R. R. Hart (Member) . Erdman (Head of Department) August 1985 ABSTRACT Improved Assessment of Population Doses and Risk Factors for a Nuclear Power Plant Under...

  20. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect (OSTI)

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01T23:59:59.000Z

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  1. Status report of advanced cladding modeling work to assess cladding performance under accident conditions

    SciTech Connect (OSTI)

    B.J. Merrill; Shannon M. Bragg-Sitton

    2013-09-01T23:59:59.000Z

    Scoping simulations performed using a severe accident code can be applied to investigate the influence of advanced materials on beyond design basis accident progression and to identify any existing code limitations. In 2012 an effort was initiated to develop a numerical capability for understanding the potential safety advantages that might be realized during severe accident conditions by replacing Zircaloy components in light water reactors (LWRs) with silicon carbide (SiC) components. To this end, a version of the MELCOR code, under development at the Sandia National Laboratories in New Mexico (SNL/NM), was modified by replacing Zircaloy for SiC in the MELCOR reactor core oxidation and material properties routines. The modified version of MELCOR was benchmarked against available experimental data to ensure that present SiC oxidation theory in air and steam were correctly implemented in the code. Additional modifications have been implemented in the code in 2013 to improve the specificity in defining components fabricated from non-standard materials. An overview of these modifications and the status of their implementation are summarized below.

  2. Technical basis for the transportation related handling representative accidents and associated hazards condition

    SciTech Connect (OSTI)

    TOMASZEWSKI, T.A.

    2003-03-21T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the handling and movement of tank farm waste sample containers, and mixed, low-level, and hazardous operational waste containers incidental to onsite vehicle transportation representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. See RPP-14286, Facility Worker Technical Basis Document, for these considerations. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described in this report.

  3. INVESTIGATION OF ACCIDENTS AND SHARING INFORMATION : AN EXPERT POINT OF VIEW

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    in explosive factories. Later, accidents were reported in flour mills (end of 18th Century), in mines (coal of plant design and Operation including human factor, management and maintenance · at competent authorities and make detailed data available. The Directory ofexisting databases prepared by ESReDA could be considered

  4. DYNAMIC ANALYSIS OF HANFORD UNIRRADIATED FUEL PACKAGE SUBJECTED TO SEQUENTIAL LATERAL LOADS IN HYPOTHETICAL ACCIDENT CONDITIONS

    SciTech Connect (OSTI)

    Wu, T

    2008-04-30T23:59:59.000Z

    Large fuel casks present challenges when evaluating their performance in the Hypothetical Accident Conditions (HAC) specified in the Code of Federal Regulations Title 10 part 71 (10CFR71). Testing is often limited by cost, difficulty in preparing test units and the limited availability of facilities which can carry out such tests. In the past, many casks were evaluated without testing by using simplified analytical methods. This paper presents a numerical technique for evaluating the dynamic responses of large fuel casks subjected to sequential HAC loading. A nonlinear dynamic analysis was performed for a Hanford Unirradiated Fuel Package (HUFP) [1] to evaluate the cumulative damage after the hypothetical accident Conditions of a 30-foot lateral drop followed by a 40-inch lateral puncture as specified in 10CFR71. The structural integrity of the containment vessel is justified based on the analytical results in comparison with the stress criteria, specified in the ASME Code, Section III, Appendix F [2], for Level D service loads. The analyzed cumulative damages caused by the sequential loading of a 30-foot lateral drop and a 40-inch lateral puncture are compared with the package test data. The analytical results are in good agreement with the test results.

  5. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    SciTech Connect (OSTI)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139-4307 (United States)

    2012-07-01T23:59:59.000Z

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO{sub 2} volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  6. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    SciTech Connect (OSTI)

    Audin, L.

    1990-12-01T23:59:59.000Z

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs.

  7. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    SciTech Connect (OSTI)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01T23:59:59.000Z

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  8. Generation IV benchmarking of TRISO fuel performance models under accident conditions. Modeling input data

    SciTech Connect (OSTI)

    Blaise Collin

    2014-09-01T23:59:59.000Z

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary.

  9. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    SciTech Connect (OSTI)

    Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

    1984-04-01T23:59:59.000Z

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

  10. Revisiting Insights from Three Mile Island Unit 2 Postaccident Examinations and Evaluations in View of the Fukushima Daiichi Accident

    SciTech Connect (OSTI)

    Joy Rempe; Mitchell Farmer; Michael Corradini; Larry Ott; Randall Gauntt; Dana Powers

    2012-11-01T23:59:59.000Z

    The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized water reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.

  11. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    SciTech Connect (OSTI)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01T23:59:59.000Z

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  12. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    SciTech Connect (OSTI)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01T23:59:59.000Z

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  13. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    SciTech Connect (OSTI)

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R. [Anatech, San Diego, CA (United States)

    1998-08-01T23:59:59.000Z

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  14. Unavoidable Accident

    E-Print Network [OSTI]

    Grady, Mark F.

    2009-01-01T23:59:59.000Z

    463. _____. 1987. Economic Analysis of Accident Law. _____.2005. “Liability for Accidents”, NBER Working Paper No.possibility is that the accident wasn’t under the defendant’

  15. Study of Air Ingress Across the Duct During the Accident Conditions

    SciTech Connect (OSTI)

    Hassan, Yassin

    2013-05-06T23:59:59.000Z

    The goal of this project is to study the fundamental physical phenoena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a nupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefor, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that mimimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlaying phenomena. The combination of interdiffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. THis project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses Perform computational fluid dynamics analysis of air ingress phenomena

  16. Analysis of containment performance and radiological consequences under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.

    1994-01-01T23:59:59.000Z

    A severe accident study was conducted to evaluate conservatively scoped source terms and radiological consequences to support the Advanced Neutron Source (ANS) Conceptual Safety Analysis Report (CSAR). Three different types of severe accident scenarios were postulated with a view of evaluating conservatively scoped source terms. The first scenario evaluates maximum possible steaming loads and associated radionuclide transport, whereas the next scenario is geared towards evaluating conservative containment loads from releases of radionuclide vapors and aerosols with associated generation of combustible gases. The third scenario follows the prescriptions given by the 10 CFR 100 guidelines. It was included in the CSAR for demonstrating site-suitability characteristics of the ANS. Various containment configurations are considered for the study of thermal-hydraulic and radiological behaviors of the ANS containment. Severe accident mitigative design features such as the use of rupture disks were accounted for. This report describes the postulated severe accident scenarios, methodology for analysis, modeling assumptions, modeling of several severe accident phenomena, and evaluation of the resulting source term and radiological consequences.

  17. FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

    SciTech Connect (OSTI)

    Rest, J.; Zawadzki, S.A. (Argonne National Lab., IL (United States))

    1992-09-01T23:59:59.000Z

    The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified.

  18. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    SciTech Connect (OSTI)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01T23:59:59.000Z

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  19. View

    E-Print Network [OSTI]

    2004-12-25T23:59:59.000Z

    Let fbe a meromorphic function satisfying condition (1.2), and let rj be a sequence with property (2.5). Then the set S is finite and for some subsequence of ...

  20. UNIVERSITY OF TRENTO ACCIDENT INSURANCE

    E-Print Network [OSTI]

    1 UNIVERSITY OF TRENTO ACCIDENT INSURANCE POLICY This document reflects the contractual conditions in force, though it should not be considered as a binding analysis of the coverage and, in case of accident for the purposes stated. TYPE OF COVERAGE = GROUP ACCIDENT INSURANCE POLICY No. = 088 00429120 COMPANY NAME

  1. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    SciTech Connect (OSTI)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01T23:59:59.000Z

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  2. Accident Investigations

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

    2011-03-04T23:59:59.000Z

    This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities.

  3. ASSESSING CAUSAL FACTORS IN INDIVIDUAL ROAD ACCIDENTS

    E-Print Network [OSTI]

    Minnesota, University of

    ASSESSING CAUSAL FACTORS IN INDIVIDUAL ROAD ACCIDENTS: COLLECTIVE RESPONSIBILITY IN FREEWAY REAR accident report: Happened on I-94 in downtown Minneapolis Happened during the afternoon peak period Vehicle" is a "condition or event" such that "had the condition or event been prevented...the accident would not occur

  4. Accident Investigations

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

    1996-04-26T23:59:59.000Z

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment, safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 2, 4-26-96

  5. Accident Investigations

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

    1995-10-26T23:59:59.000Z

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment , safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 1, 10-26-95. Cancels parts of DOE 5484.1

  6. Nuclear criticality safety tools in the Chernobyl-4 accident analysis

    SciTech Connect (OSTI)

    Landeyro, P.A.

    1988-01-01T23:59:59.000Z

    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.

  7. Employee Accident / Incident Investigation Report Employee Name _________________________________________________________________

    E-Print Network [OSTI]

    Long, Nicholas

    Employee Accident / Incident Investigation Report Employee Name's Title _________________________________________________________________ Date and Time of Accident accident occurred

  8. Accident/Injury Reporting, Investigation, & Basic First Aid Plan

    E-Print Network [OSTI]

    Long, Nicholas

    Accident/Injury Reporting, Investigation, & Basic First Aid Plan Environmental Health, Safety of accidents/injuries at Stephen F. Austin State University (SFASU) and provides basic first aid practices. It is designed to help reduce injuries by reducing unsafe or hazardous conditions and discouraging accident

  9. Estimating Pedestrian Accident Exposure: Protocol Report

    E-Print Network [OSTI]

    Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01T23:59:59.000Z

    Pedestrian Accident Risk. Accident Analysis and Prevention,Pedestrian Accidents. Accident Analysis and Prevention, Vol.in New Zealand. Accident Analysis and Prevention, Vol. 27,

  10. Placental findings in cord accidents

    E-Print Network [OSTI]

    Parast, Mana M

    2012-01-01T23:59:59.000Z

    Placental findings in cord accidents. BMC Pregnancy andPlacental findings in cord accidents Mana M Parast Fromfor stillbirth. “Cord accident,” defined by obstruction of

  11. The Accident Externality from Driving

    E-Print Network [OSTI]

    Edlin, Aaron S.; Karaca-Mandic, Pinar

    2007-01-01T23:59:59.000Z

    Sex-Divided Mile- age, Accident, and Insurance Cost DataMandic. 2003. “The Accident Externality from Driving. ”Insurance Res. Council. accident externality from driving

  12. Interpreting Accident Statistics

    E-Print Network [OSTI]

    Ferreira, Joseph Jr.

    Accident statistics have often been used to support the argument that an abnormally small proportion of drivers account for a large proportion of the accidents. This paper compares statistics developed from six-year data ...

  13. A Road Accident

    E-Print Network [OSTI]

    G.yu lha

    tracks (include description/relationship if appropriate) NA Title of track A Road Accident Translation of title Description (to be used in archive entry) Shel ko shares his experience of a serious road accident in which the truck he...

  14. Accident motivates scholarship recipient

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

    Accident motivates scholarship recipient Leyba encourages students: apply for Los Alamos Employees' Scholarship Fund Life-changing experience: springboard to a career in exercise,...

  15. The Accident Externality from Driving

    E-Print Network [OSTI]

    Edlin, Aaron S.; Karaca-Mandic, Pinar

    2005-01-01T23:59:59.000Z

    a given state could a?ect accident risk and could correlateVolume on Motor-Vehicle Accidents on Two-Lane Tangents. ”Laurie. “Sex-Divided Mileage Accident and In- surance Cost

  16. The Accident Externality from Driving

    E-Print Network [OSTI]

    Edlin, Aaron S.; Karaca-Mandic, Pinar

    2003-01-01T23:59:59.000Z

    to which this externality results from increases in accidentrates, accident severity or both remains unclear. Itpertains to underinsured accident costs like fatality risk.

  17. Radiological Release Accident Investigation Report

    Broader source: Energy.gov [DOE]

    Phase 1 of this accident investigation report is an independent product of the Accident Investigation Board appointed by Matthew Moury, Deputy Assistant Secretary, Safety, Security, and Quality...

  18. APS Guideline for Accident Investigations

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

    APS Guideline for Accident Investigations Introduction Purpose The primary purpose of an incident or accident investigation is to identify the hazard control systems that either...

  19. TIPS ON ACCIDENT/INCIDENT REPORTING Accident Reporting Why?

    E-Print Network [OSTI]

    Lennard, William N.

    TIPS ON ACCIDENT/INCIDENT REPORTING Accident Reporting ­ Why? Obligation to report Health Care of the accident ­ if not, the organization (i.e. the department) can be fined Obligation under Section 51, 52 happened? When did it happen? (Date, Time and Place) When was the accident/incident reported? Any

  20. Improving Transportation Safety Through Accident

    E-Print Network [OSTI]

    Minnesota, University of

    ;10! Investigative Groups ·" Highway Factors & Bridge Construction ·" Bridge Design ·" Witness ·" Survival accidents. ·" Major Railroad accidents. ·" Major Pipeline accidents. ·" Major marine accidents of the U10 gusset plates, due to a design error by the bridge design firm . . . Contributing to the design

  1. NORTHWESTERN UNIVERSITY ACCIDENT REPORT FORM

    E-Print Network [OSTI]

    Shahriar, Selim

    NORTHWESTERN UNIVERSITY ACCIDENT REPORT FORM Whenever a University vehicle sustains damage of any kind, or is involved in an accident which results in personal injury or property damage, this accident that this form is for University Use Only and is not meant to supersede the official state accident report form

  2. Accident Response Group

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

    1991-09-20T23:59:59.000Z

    To establish Department of Energy (DOE) policy for DOE response to accidents and significant incidents involving nuclear weapons or nuclear weapon components. Cancels DOE O 5530.1. Canceled by DOE O 153.1.

  3. Accident resistant transport container

    DOE Patents [OSTI]

    Andersen, John A. (Albuquerque, NM); Cole, James K. (Albuquerque, NM)

    1980-01-01T23:59:59.000Z

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  4. Accident Analysis and Prevention 36 (2004) 933946 Freeway safety as a function of traffic flow

    E-Print Network [OSTI]

    Detwiler, Russell

    2004-01-01T23:59:59.000Z

    Accident Analysis and Prevention 36 (2004) 933­946 Freeway safety as a function of traffic flow of strong relationships between traffic flow conditions and the likelihood of traffic accidents (crashes reserved. Keywords: Traffic safety; Accident rates; Traffic flow; Loop detectors; Speed; Traffic density

  5. Analysis of Comair flight 5191 with the Functional Resonance Accident Model

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Analysis of Comair flight 5191 with the Functional Resonance Accident Model Erik Hollnagel 1 Abstract The goal of an accident investigation is to determine why a certain combination of conditions, events, and actions led to the specific outcome. Accidents in complex high risk operations

  6. Explaining the road accident risk: weather effects Ruth Bergel-Hayat1*

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Explaining the road accident risk: weather effects Ruth Bergel-Hayat1* , Mohammed Debbarh1 conditions and road accident risk at an aggregate level and on a monthly basis, in order to improve road accidents. Time series analysis models with explanatory variables that measure the weather quantitatively

  7. Radiological Release Accident Investigation Report - Phase 1...

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

    Radiological Release Accident Investigation Report - Phase 1 Radiation Report Radiological Release Accident Investigation Report - Phase 1 Radiation Report Phase 1 of this accident...

  8. Estimating Pedestrian Accident Exposure: Protocol Report

    E-Print Network [OSTI]

    Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01T23:59:59.000Z

    A Method of Measuring Exposure to Pedestrian Accident Risk.Accident Analysis and Prevention, Vol. 14, 1982, pp 397-405.Estimating Pedestrian Accident Exposure: Protocol Report,

  9. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01T23:59:59.000Z

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  10. Accident at Creswell Colliery, Derbyshire 

    E-Print Network [OSTI]

    Bryan, Andrew

    MINISTRY OF FUEL AND POWER ACCIDENT AT CRESWELL COLLIERY, DERBYSHIRE REPORT On the causes of, and the circumstances attending, the accident which occurred at Creswell Colliery, Derbyshire, on the 26th September, 1950 BY ...

  11. Accident Report Form Victim's Name

    E-Print Network [OSTI]

    Amin, S. Massoud

    Accident Report Form Date: Victim's Name: Address: Classification: Program Area: Activity: Brief Description of Accident: Body Fluid Spill: Action Taken by DRS Employee: Witness Name: Witness Address:____________________________________ DOB: Intramurals Front Back Revision - 2011 **Location of Accident** URC North Gymnasium URC South

  12. Electrical shock accident investigation

    SciTech Connect (OSTI)

    Not Available

    1994-09-30T23:59:59.000Z

    This report documents results of the accident investigation of an electrical shock received by two subcontractor employees on May 13, 1994, at the Pinellas Plant. The direct cause of the electrical shock was worker contact with a cut ``hot`` wire and a grounded panelboard (PPA) enclosure. Workers presumed that all wires in the enclosure were dead at the time of the accident and did not perform thorough Lockout/Tagout (LO/TO). Three contributing causes were identified. First, lack of guidance in the drawing for the modification performed in 1987 allowed the PPA panel to be used as a junction box. The second contributing cause is that Environmental, Safety and Health (ES&H) procedures do not address multiple electrical sources in an enclosure. Finally, the workers did not consider the possibility of multiple electrical sources. The root cause of the electrical shock was the inadequacy of administrative controls, including construction requirement and LO/TO requirements, and subcontractor awareness regarding multiple electrical sources. Recommendations to prevent further reoccurrence of this type of accident include revision of ES&H Standard 2.00, Electrical Safety Program Manual, to document requirements for multiple electrical sources in a single enclosure to specify a thorough visual inspection as part of the voltage check process. In addition, the formality of LO/TO awareness training for subcontractor electricians should be increased.

  13. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    SciTech Connect (OSTI)

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01T23:59:59.000Z

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  14. Accident Investigation of the June 17, 2012, Construction Accident...

    Energy Savers [EERE]

    June 17, 2012, Construction Accident - Structural Steel Collapse at The Over pack Storage Expansion 2 at the Naval Reactors Facility at the Idaho National Laboratory, Idaho Falls,...

  15. Evaluating the effectiveness of wildlife accident mitigation installations with the wildlife accident reporting system (WARS) in British Columbia

    E-Print Network [OSTI]

    Sielecki, Leonard E.

    2001-01-01T23:59:59.000Z

    EFFECTIVENESS OF WILDLIFE ACCIDENT MITIGATION INSTALLATIONSWITH THE WILDLIFE ACCIDENT REPORTING SYSTEM (WARS) INadministers the Wildlife Accident Reporting System (WARS), a

  16. Accident motivates scholarship recipient

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting the TWP TWP Related LinksATHENA couldAboutClean WaterAccessingAccident

  17. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

    2014-10-01T23:59:59.000Z

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.

  18. EPR Severe Accident Threats and Mitigation

    SciTech Connect (OSTI)

    Azarian, G. [Framatome ANP SAS, Tour Areva, Place de la Coupole 92084 Paris la Defense (France); Kursawe, H.M.; Nie, M.; Fischer, M.; Eyink, J. [Framatome ANP GmbH, Freyeslebenstrasse, 1, D-91058 Erlangen (Germany); Stoudt, R.H. [Framatome ANP Inc. - 3315 Old Forest Rd, Lynchburgh, VA 24501 (United States)

    2004-07-01T23:59:59.000Z

    Despite the extremely low EPR core melt frequency, an improved defence-in-depth approach is applied in order to comply with the EPR safety target: no stringent countermeasures should be necessary outside the immediate plant vicinity like evacuation, relocation or food control other than the first harvest in case of a severe accident. Design provisions eliminate energetic events and maintain the containment integrity and leak-tightness during the entire course of the accident. Based on scenarios that cover a broad range of physical phenomena and which provide a sound envelope of boundary conditions associated with each containment challenge, a selection of representative loads has been done, for which mitigation measures have to cope with. This paper presents the main critical threats and the approach used to mitigate those threats. (authors)

  19. Probability of spent fuel transportation accidents

    SciTech Connect (OSTI)

    McClure, J. D.

    1981-07-01T23:59:59.000Z

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10/sup -7/ spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10/sup -9//mile.

  20. BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE

    E-Print Network [OSTI]

    Suzuki, Masatsugu

    BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE Especially Designed for Students of insurance. Your coverage is governed by a policy of student accident and sickness insurance underwritten

  1. BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE

    E-Print Network [OSTI]

    Suzuki, Masatsugu

    BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE Especially Designed for International Students is governed by a policy of student accident and sickness insurance underwritten by BCS Insurance Company BCS

  2. OSSA - An optimized approach to severe accident management: EPR application

    SciTech Connect (OSTI)

    Sauvage, E. C.; Prior, R.; Coffey, K. [AREVA, FRAMATOME-ANP SAS, Paris, 92084 La Defense (France); Mazurkiewicz, S. M. [AREVA, FRAMATOME-ANP Inc, Lynchburg, VA 24506-0935 (United States)

    2006-07-01T23:59:59.000Z

    There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field. This revised approach will incorporate a number of new features which will simplify and streamline the guidance material while ensuring comprehensive guidance for response to any severe accident. Examples of such features include : - Identification of severe accident challenges based on plant specific studies. - Revision of the split of responsibilities between operations and technical support center staff. - Fixed setpoint entry conditions, ensuring that the transition from emergency procedures takes place at a consistent core/fuel condition (regardless of scenario), and which fixes the time window available to attempt ultimate preventive measures. - A safety function concept for monitoring plant conditions (in the control room). - An integrated graphic-based diagnostic tool including entry condition, challenge prioritization, and exit condition monitoring to be used by the technical support team. This paper describes the basic features of OSSA, and project status. (authors)

  3. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect (OSTI)

    Su'ud, Zaki; Anshari, Rio [Nuclear and Biophysics Research Group, Dept. of Physics, Bandung Institute of Technology, Jl.Ganesha 10, Bandung, 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  4. A review of criticality accidents

    SciTech Connect (OSTI)

    Stratton, W R; Smith, D R

    1989-03-01T23:59:59.000Z

    Criticality accidents and the characteristics of prompt power excursions are discussed. Forty-one accidental power transients are reviewed. In each case where available, enough detail is given to help visualize the physical situation, the cause or causes of the accident, the history and characteristics of the transient, the energy release, and the consequences, if any, to personnel and property. Excursions associated with large power reactors are not included in this study, except that some information on the major accident at the Chernobyl reactor in April 1986 is provided in the Appendix. 67 refs., 21 figs., 2 tabs.

  5. Implications for accident management of adding water to a degrading reactor core

    SciTech Connect (OSTI)

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-02-01T23:59:59.000Z

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  6. LOCA and Air Ingress Accident Analysis of a Pebble Bed Reactor

    E-Print Network [OSTI]

    1 LOCA and Air Ingress Accident Analysis of a Pebble Bed Reactor by Tieliang Zhai Submitted Accident Analysis of a Pebble Bed Reactor by Tieliang Zhai Submitted to the Department of Nuclear features of the pebble bed reactor under challenging conditions. The first part of the thesis explored

  7. University of Pittsburgh Vehicle Accident Report Form

    E-Print Network [OSTI]

    Sibille, Etienne

    University of Pittsburgh Vehicle Accident Report Form To be completed by the driver immediately following the accident (if medically able) and return this completed form to Fleet Services, Dept of Parking-624-1817 A. Report Date: ______/______/_______ B: Accident Data Date of accident

  8. Exact Location : Date of Accident : AM PM

    E-Print Network [OSTI]

    Swaddle, John

    SSN Cell Phone Home Phone Work Phone Exact Location : Date of Accident : AM PM Date accident treatment provided? Yes No Where Was time lost from work? Yes No If yes, how long? Could this accident have the following information as soon as it relates to your work related accident/injury/illness within 72 hours

  9. Bordereau de transmission accident du travail

    E-Print Network [OSTI]

    Pouyanne, Nicolas

    Bordereau de transmission accident du travail Service des pensions et accidents du travail accidents du travail du CNRS Accompagné des pièces requises Nom .................................................... Prénom ........................ Matricule ...... Composition du dossier Observations Déclaration d'accident

  10. An Interview About an Accident

    E-Print Network [OSTI]

    G.yu lha

    2009-12-17T23:59:59.000Z

    Length of track 0:02:32 Related tracks (include description/relationship if appropriate) Title of track An Interview About an Accident Translation of title Description (to be used in archive entry) The respondent recalls how he and his... wife survived a motorcycle accident. Genre or type (i.e. epic, song, ritual) Interview Name of recorder (if different from collector) G.yu lha Date of recording December 17th 2009 Place of recording Siyuewu Village, Puxi Township, Rangtang...

  11. Accident tolerant fuel analysis

    SciTech Connect (OSTI)

    Smith, Curtis [Idaho National Laboratory; Chichester, Heather [Idaho National Laboratory; Johns, Jesse [Texas A& M University; Teague, Melissa [Idaho National Laboratory; Tonks, Michael Idaho National Laboratory; Youngblood, Robert [Idaho National Laboratory

    2014-09-01T23:59:59.000Z

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  12. Accident Tolerant Fuel Analysis

    SciTech Connect (OSTI)

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01T23:59:59.000Z

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  13. Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

    SciTech Connect (OSTI)

    Yulianti, Yanti [Dept. of Physics, Universitas Lampung (UNILA), Jl. Sumantri Brojonegor No.1 Bandar Lampung (Indonesia); Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Su'ud, Zaki; Waris, Abdul; Khotimah, S. N. [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Shafii, M. Ali [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Dept. of Physics, Universitas Andalas (UNAND), Kampus Limau Manis, Padang, Sumatera Barat (Indonesia)

    2010-12-23T23:59:59.000Z

    The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

  14. The Wildlife Accident Reporting System (WARS) in British Columbia

    E-Print Network [OSTI]

    Sielecki, Leonard E.

    2003-01-01T23:59:59.000Z

    2001, WARS 2000 Wildlife Accident Reporting System (2000related motor vehicle accident claim data and funding toTHE WILDLIFE ACCIDENT REPORTING SYSTEM (WARS) IN BRITISH

  15. Type B Accident Investigation Board Report on the September 7...

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

    Accident Investigation Board Report on the September 7, 2001, Burn Accident at Oak Ridge National Laboratory, Building 9210 Type B Accident Investigation Board Report on the...

  16. Type B Accident Investigation of the Subcontractor Employee Injuries...

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

    Type B Accident Investigation of the Subcontractor Employee Injuries from a November 15, 2000, Fall Accident at the Oak Ridge National Laboratory Type B Accident Investigation of...

  17. Estimating Pedestrian Accident Exposure: Automated Pedestrian Counting Devices Report

    E-Print Network [OSTI]

    Bu, Fanping; Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01T23:59:59.000Z

    291. Estimating Pedestrian Accident Exposure: Draft ProtocolEstimating Pedestrian Accident Exposure: Draft Protocol39. Estimating Pedestrian Accident Exposure: Draft Protocol

  18. Preliminary Assessment of ICRP Dose Conversion Factor Recommendations for Accident Analysis Applications

    SciTech Connect (OSTI)

    Vincent, A.M.

    2002-03-13T23:59:59.000Z

    Accident analysis for U.S. Department of Energy (DOE) nuclear facilities is an integral part of the overall safety basis developed by the contractor to demonstrate facility operation can be conducted safely. An appropriate documented safety analysis for a facility discusses accident phenomenology, quantifies source terms arising from postulated process upset conditions, and applies a standardized, internationally-recognized database of dose conversion factors (DCFs) to evaluate radiological conditions to offsite receptors.

  19. accidents: Topics by E-print Network

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

    Phone Home Phone Work Phone Exact Location : Date of Accident : AM PM Date accident treatment provided? Yes No Where Was time lost from work? Yes No If yes, how long? Could this...

  20. BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE

    E-Print Network [OSTI]

    Suzuki, Masatsugu

    BLANKET STUDENT ACCIDENT AND SICKNESS INSURANCE Especially Designed for the Dependents. It is not a contract of insurance. Your coverage is governed by a policy of student accident and sickness insurance

  1. RICE UNIVERSITY ACCIDENT/INJURY REPORT

    E-Print Network [OSTI]

    Natelson, Douglas

    RICE UNIVERSITY ACCIDENT/INJURY REPORT Please Print Section A: Details of incident Injury Work Exposure to radiation Mental stress factors Noise Insect/animal bite Vehicle accident Slip

  2. Analysis of accidents during flashing operations

    E-Print Network [OSTI]

    Obermeyer, Michael Edward

    1993-01-01T23:59:59.000Z

    University, 1976 Federal Highway Administration Study, 1980 San Francisco Study National Study Portland, Oregon Study Summary of Literature Review Studies 13 14 16 17 20 CHAPTER Page III. ACCIDENT ANALYSIS METHODOLOGY . 22 Study Site Location... V. SUMMARY AND FINDINGS 44 REFERENCES 48 VITA 50 LIST OF TABLES TABLE 1. Groupings for Marson's Accident Analysis 2. Groupings for San Francisco Accident Analysis 3. Groupings for Portland Accident Analysis 4. Sample Sizes by Volume Ratio 5...

  3. Technical basis document for the evaporator dump accident

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-03-22T23:59:59.000Z

    This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for the evaporator dump representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  4. UWO Vehicle ACCIDENT REPORTING FORM

    E-Print Network [OSTI]

    Sinnamon, Gordon J.

    UWO Vehicle ­ ACCIDENT REPORTING FORM To be completed at the scene. (Important: Do not admit liability or discuss any settlement.) If there are personal injuries or severe damage to the vehicle, call 911. If vehicle is drivable and if it's safe to do so, pull to the side of road away from traffic. Put

  5. TREE FAILURES AND ACCIDENTS IN

    E-Print Network [OSTI]

    Standiford, Richard B.

    .DEPARTMENT O F AGRICULTURE GENERAL TECHNICAL REPORT PSW- 24 #12;TREE FAILURES AND ACCIDENTS IN RECREATION are major concerns. Injuries, fatalities, and high property losses occur each year as a result of tree losses associated with public occupancy. Hazard reduction can limit such losses to predefined levels

  6. The Hartford Life and Accident Insurance

    E-Print Network [OSTI]

    The Hartford Life and Accident Insurance Company Group Numbers Basic Term Life - 677984 Basic by The Hartford Life and Accident Insurance Company. (Referred to as The Hartford or Hartford.) General from an accident, the benefit will be equal to $140,000 ($70,000 basic group term life PLUS $70

  7. Most Viewed Documents for Fission And Nuclear Technologies: September...

    Office of Scientific and Technical Information (OSTI)

    under accident conditions Bomelburg, H.J. (1977) 71 Behavior of spent nuclear fuel in water pool storage Johnson, A.B. Jr. (1977) 68 Stress analysis and evaluation of a...

  8. accident conditions lessons: Topics by E-print Network

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

    set of in-class tutorials experience and classroom observations, and present several guidelines for tutorial development Colorado at Boulder, University of 249 Lessons Learned from...

  9. accident containment conditions: Topics by E-print Network

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

    Summary: of stainless steel container materials is a potential problem for long-term radioactive waste storage-to-failure of relevant stainless steels in the annealed...

  10. accident conditions comparison: Topics by E-print Network

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

    a discussion on the choice and matching of different types of Waste Water Resource Heat Pump (WWRHP) heating and air... Zhang, C.; Wang, S.; Chen, H.; Shi, Y. 2006-01-01 24...

  11. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    SciTech Connect (OSTI)

    S.O. Bader

    1999-10-18T23:59:59.000Z

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.

  12. Computerized Accident Incident Reporting System | Department...

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

    and other accidents that occur during DOE operations. CAIRS is a Government computer system and, as such, has security requirements that must be followed. Access to the...

  13. ORISE: REAC/TS Radiation Accident Registries

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

    Accident Registries The Radiation Emergency Assistance CenterTraining Site (REACTS) at the Oak Ridge Institute for Science and Education (ORISE) maintains a number of radiation...

  14. DOE Accident Prevention and Investigation Program | Department...

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

    tools utilized in the investigation of "accidents" can be valuable in looking at leading indicators associated with our safety program, to determine the embedded precursors to...

  15. Type B Accident Investigation, Subcontractor Employee Personal...

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

    February 18, 2003, at the East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation, Subcontractor Employee Personal Protective Equipment Ignition Incident...

  16. Date of Accident: _____/_____/________ Day of Week: __________________ Hour: _____:______ AM / PM TIME VEHICLE ACCIDENT REPORT

    E-Print Network [OSTI]

    Farritor, Shane

    Page 1/2 Date of Accident: _____/_____/________ Day of Week: __________________ Hour: _____:______ AM / PM TIME VEHICLE ACCIDENT REPORT TO BE USED BY ALL STATE AGENCIES to make immediate report of all motor vehicle accidents involving State employees, vehicles, equipment or where highways could result

  17. Environmental Conditions Environmental Conditions

    E-Print Network [OSTI]

    Environmental Conditions Environmental Conditions Appendix II The unique geology, hydrology and instream habitat. This chapter examines how environmental conditions in the Deschutes watershed affect, the discussion characterizes the environmental conditions within three watershed areas: the Lower Deschutes

  18. Severe accident research in Canada

    SciTech Connect (OSTI)

    Simpson, L.A. [AECL Research, Pinawa, Manitoba (Canada)

    1994-12-31T23:59:59.000Z

    The reactor safety research program in Canada not only recognizes the unique features of the CANDU reactor, but is supplemented by a strong interaction with the LWR research community. This is especially so in the area of severe accidents. We participate in international programs such as Phebus FP and CSARP to take advantage of cooperative efforts on phenomena that are generic to all reactors, but also have our distinct programs in Canada on severe fuel damage, fission product chemistry, aerosol behaviour and hydrogen combustion and mitigation. These programs address the characteristics of Canadian nuclear fuel and containment design, and our own series of severe accident scenarios. The scope of the R&D encompasses separate effects experiments, model development and code development, leading to validation testing in several large integral test facilities including the Radioiodine Test Facility and the Blowdown Test Facility in the NRU reactor. We also have extensive hydrogen combustion test facilities including the Large Scale Vented Combustion Test Facility now under construction. The essence of the program is described with examples from recent experiments and analysis.

  19. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect (OSTI)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01T23:59:59.000Z

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to automatically respond to the change in reactor conditions and to result in a benign response to these events. This approach has the advantage of being relatively simple to implement, and does not face the issue of reliability since only fundamental physical phenomena are used in a passive manner, not active engineered systems. However, the challenge is to present a convincing case that such passive means can be implemented and used. The purpose of this paper is to describe this third approach in detail, the technical basis and experimental validation for the approach, and the resulting reactor performance that can be achieved for ATWS events.

  20. The Nevada railroad system: Physical, operational, and accident characteristics

    SciTech Connect (OSTI)

    NONE

    1991-09-01T23:59:59.000Z

    This report provides a description of the operational and physical characteristics of the Nevada railroad system. To understand the dynamics of the rail system, one must consider the system`s physical characteristics, routing, uses, interactions with other systems, and unique operational characteristics, if any. This report is presented in two parts. The first part is a narrative description of all mainlines and major branchlines of the Nevada railroad system. Each Nevada rail route is described, including the route`s physical characteristics, traffic type and volume, track conditions, and history. The second part of this study provides a more detailed analysis of Nevada railroad accident characteristics than was presented in the Preliminary Nevada Transportation Accident Characterization Study (DOE, 1990).

  1. HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND

    E-Print Network [OSTI]

    Fabrikant, J.I.

    2010-01-01T23:59:59.000Z

    Commission on the Accident at Three Mile Island (Fabrikant,Commission on the Accident at Three Mile Island. (Fahrikant,Commission on the Accident at Three Mile Island. (Fabrikant,

  2. accident victims: Topics by E-print Network

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

    Every year, traffic congestion and traffic accidents have been Cho, Sung-Bae 118 The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson,...

  3. accident zone osobennosti: Topics by E-print Network

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

    Every year, traffic congestion and traffic accidents have been Cho, Sung-Bae 62 The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson,...

  4. accident victim conduite: Topics by E-print Network

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

    Every year, traffic congestion and traffic accidents have been Cho, Sung-Bae 152 The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson,...

  5. Insights into the behavior of nuclear power plant containments during severe accidents

    SciTech Connect (OSTI)

    Horschel, D.S.; Ludwigsen, J.S.; Parks, M.B.; Lambert, L.D. [Sandia National Labs., Albuquerque, NM (United States); Dameron, R.A.; Rashid, Y.R. [ANATECH Research Corp., San Diego, CA (United States)

    1993-06-01T23:59:59.000Z

    The containment building surrounding a nuclear reactor offers the last barrier to the release of radioactive materials from a severe accident into the environment. The loading environment of the containment under severe accident conditions may include much greater than design pressures and temperatures. Investigations into the performance of containments subject to ultimate or failure pressure and temperature conditions have been performed over the last several years through a program administered by the Nuclear Regulatory Commission (NRC). These NRC sponsored investigations are subsequently discussed. Reviewed are the results of large scale experiments on reinforced concrete, prestressed concrete, and steel containment models pressurized to failure. In conjunction with these major tests, the results of separate effect testing on many of the critical containment components; that is, aged and unaged seals, a personnel air lock and electrical penetration assemblies subjected to elevated temperature and pressure have been performed. An objective of the NRC program is to gain an understanding of the behavior of typical existing and planned containment designs subject to postulated severe accident conditions. This understanding has led to the development of experimentally verified analytical tools that can be applied to accurately predict their ultimate capacities useful in developing severe accident mitigation schemes. Finally, speculation on the response of containments subjected to severe accident conditions is presented.

  6. MELCOR accident analysis for ARIES-ACT

    E-Print Network [OSTI]

    California at San Diego, University of

    Flow Flow #12;Fusion Safety Program · MELCOR is a code originally designed to model severe accidentMELCOR accident analysis for ARIES-ACT Paul Humrickhouse Brad Merrill INL Fusion Safety Program progression in water-cooled fission reactors · INL has modified it for fusion; MELCOR 1.8.5 for fusion has

  7. Does Daylight Savings Time Affect Traffic Accidents?

    E-Print Network [OSTI]

    Deen, Sophia 1988-

    2012-04-20T23:59:59.000Z

    This paper studies the effect of changes in accident pattern due to Daylight Savings Time (DST). The extension of the DST in 2007 provides a natural experiment to determine whether the number of traffic accidents is affected by shifts in hours...

  8. TRAVEL ACCIDENT INSURANCE PLAN 01-01-2012 The Travel Accident Insurance Plan provides 24-hour Accident coverage while on Authorized

    E-Print Network [OSTI]

    Johnson, Peter D.

    1 TRAVEL ACCIDENT INSURANCE PLAN 01-01-2012 The Travel Accident Insurance Plan provides 24-hour Accident coverage while on Authorized Business Travel. Coverage begins at the actual starting point. Please note that the Employer reserves the right to amend or terminate this Travel Accident Insurance

  9. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect (OSTI)

    Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

    2012-11-01T23:59:59.000Z

    This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

  10. Commercial SNF Accident Release Fractions

    SciTech Connect (OSTI)

    J. Schulz

    2004-11-05T23:59:59.000Z

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the container that confines the fuel assemblies could provide an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. This analysis, however, does not take credit for the additional barrier and establishes only the total release fractions for bare unconfined intact commercial SNF assemblies, which may be conservatively applied to confined intact commercial I SNF assemblies.

  11. Type B Accident Investigation Board Report on the November 1...

    Office of Environmental Management (EM)

    B Accident Investigation Board Report on the November 1, 1999, Construction Injury at the Monticello Mill Tailings Remedial Action Site, Monticello, Utah Type B Accident...

  12. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Savers [EERE]

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to...

  13. ORISE: The Medical Basis for Radiation-Accident Preparedness...

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

    The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REACTS Symposium on the Medical Basis for Radiation-Accident...

  14. Type A Accident Investigation of the June 21, 2001, Drilling...

    Office of Environmental Management (EM)

    A Accident Investigation of the June 21, 2001, Drilling Rig Operator Injury at the Fermi National Accelerator Laboratory, August 2001 Type A Accident Investigation of the June 21,...

  15. accident analysis structural: Topics by E-print Network

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

    provides a theoretical foundation for the introduction of unique new types of accident analysis, hazard analysis, accident prevention strategies including new approaches to...

  16. accident prone drivers: Topics by E-print Network

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

    provides a theoretical foundation for the introduction of unique new types of accident analysis, hazard analysis, accident prevention strategies including new approaches to...

  17. accident related release: Topics by E-print Network

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

    age on auto accidents is examined by employing an interrupted time series analysis of monthly accident data covering the period January, 1969, through September 1973. The data ......

  18. Radiological Release Accident Investigation Report- Phase 1 Radiation Report

    Broader source: Energy.gov [DOE]

    Phase 1 of this accident investigation report is an independent product of the Accident Investigation Board appointed by Matthew Moury, Deputy Assistant Secretary, Safety, Security, and Quality...

  19. Type B Accident Investigation of the July 12, 2007, Forklift...

    Energy Savers [EERE]

    2, 2007, Forklift and Pedestrian Accident at the Paducah Gaseous Diffusion Plant, PortsmouthPaducah Project Office Type B Accident Investigation of the July 12, 2007, Forklift and...

  20. Type B Accident Investigation on the February 17, 2004, Personal...

    Energy Savers [EERE]

    Investigation of the July 12, 2007, Forklift and Pedestrian Accident at the Paducah Gaseous Diffusion Plant, PortsmouthPaducah Project Office Type B Accident Investigation...

  1. Web Based Course: SAF-230DE, Accident Investigation Overview...

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

    Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video September 20, 2013 -...

  2. Partnership Logging Accidents Cornelis de Hoop, LA Forest Products Lab

    E-Print Network [OSTI]

    Partnership Logging Accidents · by · Cornelis de Hoop, LA Forest Products Lab · Albert Lefort Agreement · 1998 & 1999 Accident Reports · 25 injuries reported · 185 loggers signed up · 8 deaths 1999

  3. accident management aids: Topics by E-print Network

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

    Accident, Illness and Liability Coverage Risk Management in the 4-H Youth Development Program Environmental Sciences and Ecology Websites Summary: 1 Accident, Illness and...

  4. Accident Investigation of the June 1, 2013, Stairway Fall Resulting...

    Energy Savers [EERE]

    Accident Investigation of the June 1, 2013, Stairway Fall Resulting in a Federal Employee Fatality at DOE Headquarters Germantown, Maryland Accident Investigation of the June 1,...

  5. accident sequence precursor: Topics by E-print Network

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

    Susskind; Nicolaos Toumbas 2000-03-17 7 Pedestrian Accidents - In-depth Analysis and Accident Figures. Open Access Theses and Dissertations Summary: ?? Pedestrian fatalities and...

  6. accident phenomenology cours: Topics by E-print Network

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

    Demande Commentaires Parrott, Lael 211 Pedestrian Accidents - In-depth Analysis and Accident Figures. Open Access Theses and Dissertations Summary: ?? Pedestrian fatalities and...

  7. accident management summary: Topics by E-print Network

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

    Management of the Acute Radiation Syndrome 2001 flow Feed back Radiation Accident MedicalManagement COMPENDIUMCOMPENDIUM MEDICAL MANAGEMENT OF RADIATION ACCIDENTS...

  8. Overview of the U.S. DOE Accident Tolerant Fuel Development Program

    SciTech Connect (OSTI)

    Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton; Lance L. Snead

    2013-09-01T23:59:59.000Z

    The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.

  9. Prarie View RDF

    Energy Savers [EERE]

    PRAIRIE VIEW RDF 2 Prairie View RDF Located at JAAP (approx. 40 miles southwest of Chicago), 223 acres on 455 Acre Parcel Will County Owner; Waste Management, Operator ...

  10. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

    Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

    2004-05-15T23:59:59.000Z

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  11. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01T23:59:59.000Z

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  12. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01T23:59:59.000Z

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  13. Estimating Rear-End Accident Probabilities at Signalized Intersections: An Occurrence-Mechanism Approach

    E-Print Network [OSTI]

    Wang, Yinhai

    Estimating Rear-End Accident Probabilities at Signalized Intersections: An Occurrence intersections, rear-end accidents are frequently the predominant accident type. These accidents result from to this deceleration. This paper mathematically represents this process, by expressing accident probability

  14. Markov Model of Severe Accident Progression and Management

    SciTech Connect (OSTI)

    Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

    2012-06-25T23:59:59.000Z

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  15. Site restoration: Estimation of attributable costs from plutonium-dispersal accidents

    SciTech Connect (OSTI)

    Chanin, D.I.; Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)

    1996-05-01T23:59:59.000Z

    A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.

  16. COMPARING THE IDENTIFICATION OF RECOMMENDATIONS BY DIFFERENT ACCIDENT

    E-Print Network [OSTI]

    Johnson, Chris

    will be identified for similar incidents. Accident analysis methods can also help to reduce individual bias

  17. SUPERVISOR'S ACCIDENT INVESTIGATION FORM Employee's Name: Job Title

    E-Print Network [OSTI]

    Jiang, Wen

    SUPERVISOR'S ACCIDENT INVESTIGATION FORM Employee's Name: Job Title: Time employee has been in current position? How long had employee been at work prior to injury? Accident Date: Time of Accident: AM PM Overtime: Yes No Location of Accident (Be Specific): Specific Task Being Performed at Time

  18. ACCIDENT PREVENTION SIGNS, TAGS, LABELS, SIGNALS, PIPING SYSTEM IDENTIFICATION AND

    E-Print Network [OSTI]

    US Army Corps of Engineers

    EM 385-1-1 XX Sep 13 i Section 8 ACCIDENT PREVENTION SIGNS, TAGS, LABELS, SIGNALS, PIPING SYSTEM............................................................8-13 Tables: 8-1 Accident Prevention Sign Requirements..........................8-17 8-2 Accident.......................................8-24 8-9 Accident Prevention Tags.............................................8-25 #12;EM 385-1-1 XX

  19. STATE OF CALIFORNIA -DGS ORIM VEHICLE ACCIDENT REPORT

    E-Print Network [OSTI]

    Ponce, V. Miguel

    STATE OF CALIFORNIA - DGS ORIM VEHICLE ACCIDENT REPORT STD. 270 (REV. 2/2002c) ACCIDENT PREVIOUSLY REPORTED TO ORIM? (If Yes, give date) YES NO THIS REPORT MUST BE MAILED WITHIN 48 HOURS AFTER ACCIDENT (ACCIDENTS INVOLVING INJURY SHOULD FIRST BE CALLED OR FAXED TO ORIM AT (916) 376-5302 - CALNET 480-5302 - FAX

  20. UoS Motor Accident Report Form COMPANY DETAILS

    E-Print Network [OSTI]

    Sussex, University of

    UNIV01FL02 UoS Motor Accident Report Form COMPANY DETAILS INSURED: University of Sussex ADDRESS: LOCATION: DESCRIPTION OF HOW ACCIDENT HAPPENED: PLEASE DRAW A SKETCH OF THE ACCIDENT: #12;DRIVER DETAILS: PREVIOUS ACCIDENTS: ADDRESS: VEHICLE DETAILS DATE VEHICLE PURCHASED: MAKE/MODEL: REGISTRATION: MILEAGE

  1. Review of models applicable to accident aerosols

    SciTech Connect (OSTI)

    Glissmeyer, J.A.

    1983-07-01T23:59:59.000Z

    Estimations of potential airborne-particle releases are essential in safety assessments of nuclear-fuel facilities. This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident-generated aerosol sources. Such characterization of the accident-generated aerosols is a necessary step toward estimating their eventual release in any accident scenario. Existing aerosol models can predict the size distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation, and other phenomena. Models developed in the fields of fluid mechanics, indoor air pollution, and nuclear-reactor accidents are reviewed with this nuclear fuel facility application in mind. The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity.

  2. Dose calculations for severe LWR accident scenarios

    SciTech Connect (OSTI)

    Margulies, T.S.; Martin, J.A. Jr.

    1984-05-01T23:59:59.000Z

    This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well.

  3. Use of root in vehicular accident reconstruction

    E-Print Network [OSTI]

    Scurlock, Bob

    2011-01-01T23:59:59.000Z

    The purpose of this article is to introduce the reader to the ROOT data analysis software package, and demonstrate how it may be used to complement one's accident reconstruction analyses.

  4. A systems approach to food accident analysis

    E-Print Network [OSTI]

    Helferich, John D

    2011-01-01T23:59:59.000Z

    Food borne illnesses lead to 3000 deaths per year in the United States. Some industries, such as aviation, have made great strides increasing safety through careful accident analysis leading to changes in industry practices. ...

  5. An analysis of evacuation options for nuclear accidents

    SciTech Connect (OSTI)

    Tawil, J.J.; Strenge, D.L.; Schultz, R.W. [Battelle Memorial Inst., Richland, WA (United States)

    1987-11-01T23:59:59.000Z

    In this report we consider the threat posed by the accidental release of radionuclides from a nuclear power plant. The objective is to establish relationships between radiation dose and the cost of evacuation under a wide variety of conditions. The dose can almost always be reduced by evacuating the population from a larger area. However, extending the evacuation zone outward will cause evacuation costs to increase. The purpose of this analysis was to provide the Environmental Protection Agency (EPA) a data base for evaluating whether implementation costs and risks averted could be used to justify evacuation at lower doses. The procedures used and results of these analyses are being made available as background information for use by others. We develop cost/dose relationships for 54 scenarios that are based upon the severity of the reactor accident, meteorological conditions during the release of radionuclides into the environment, and the angular width of the evacuation zone. The 54 scenarios are derived from combinations of three accident severity levels, six meteorological conditions and evacuation zone widths of 70{degree}, 90{degree}, and 180{degree}.

  6. Assessing economic consequences of radiation accidents

    SciTech Connect (OSTI)

    Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

    1987-01-01T23:59:59.000Z

    This project reviewed the literature on the economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the high-level radioactive waste (HLW) disposal system before closure; determined needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system; and gathered data that might be useful for the needed revisions. 8 refs., 1 tab.

  7. risk_policies_accident_std_vist.doc/ac 1 Revised 07.26.13 STUDENT AND VISITOR ACCIDENT

    E-Print Network [OSTI]

    Su, Xiao

    risk_policies_accident_std_vist.doc/ac 1 Revised 07.26.13 STUDENT AND VISITOR ACCIDENT REPORTING: 408-924-1892 Student and Visitor Accident Reporting Guidelines These guidelines provide instructions for reporting and handling accidents or incidents that happen to students and visitors while on the San José

  8. Accident Procedure Outline the procedures for accidents involving University of Michigan (U-M) vehicles.

    E-Print Network [OSTI]

    Kirschner, Denise

    owned by U-M are covered by the U-M self insurance program administered by Risk Management. Procedure 1. An accident is defined as any incident that causes damage to people or property. 2. In the event. 4. If the accident causes personal injury to the driver, occupants and/or pedestrian, contact Risk

  9. LOCA with consequential or delayed LOOP accidents: Unique issues, plant vulnerability, and CDF contributions

    SciTech Connect (OSTI)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-08-01T23:59:59.000Z

    A loss-of-coolant accident (LOCA) can cause a loss-of-offsite power (LOOP) wherein the LOOP is usually delayed by few seconds or longer. Such an accident is called LOCA with consequential LOOP, or LOCA with delayed LOOP (here, abbreviated as LOCA/LOOP). This paper analyzes the unique conditions that are associated with a LOCA/LOOP, presents a model, and quantifies its contribution to core damage frequency (CDF). The results show that the CDF contribution can be a dominant contributor to risk for certain plant designs, although boiling water reactors (BWRs) are less vulnerable than pressurized water reactors (PWRs).

  10. Creating an urban deer-vehicle accident management plan using information from a town's GIS project

    E-Print Network [OSTI]

    Premo, Dean B.; Rogers, Elizabeth I.

    2001-01-01T23:59:59.000Z

    AN URBAN DEER-VEHICLE ACCIDENT MANAGEMENT PLAN USINGincrease in deer vehicle accidents. Given the Town'sof increased deer vehicle accidents which, in the past 10

  11. Do "Accidents" Happen? An Examination of Injury Mortality Among Maltreated Children

    E-Print Network [OSTI]

    Hornstein, Emily Putnam

    2010-01-01T23:59:59.000Z

    2002;26. Garling T. Children's environments, accidents,and accident prevention: An introduction. In: Garling T,Toward a Psychology of Accident Prevention. New York: Plenum

  12. A research university's rapid response to a fatal chemistry accident: Safety changes and outcomes

    E-Print Network [OSTI]

    Gibson, JH; Schröder, I; Wayne, NL

    2014-01-01T23:59:59.000Z

    to a fatal chemistry accident: Safety changes and outcomesprogram following a chemistry accident in December 2008 thatcommunity. Since the 2008 accident at UCLA, the na- tional

  13. Exploratory Analysis of Motor Carrier Accident Risk And Daily Driving Patterns

    E-Print Network [OSTI]

    Jovanis, Paul P.; Kaneko, Tetsuya; Lin, Tzuoo-Din

    1991-01-01T23:59:59.000Z

    in a Sleeper Berth," Accident Analysis and Prevention. 1988,Survival Theory," Accident Analysis and Prevention, Vol. 21,most of the analyses with accident data compared actual

  14. Estimating Pedestrian Accident Exposure: Approaches to a Statewide Pedestrian Exposure Database

    E-Print Network [OSTI]

    Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01T23:59:59.000Z

    in New Zealand. Accident Analysis and Prevention, Vol. 27,System Network-Traffic Accident Analysis and SurveillanceAutomated Traffic Accident Surveillance and Analysis System,

  15. Multiday Driving Patterns and Motor Carrier Accident Risk: A Disagregate Analysis

    E-Print Network [OSTI]

    Kaneko, Tetsuya; Jovanis, Paul P.

    1991-01-01T23:59:59.000Z

    as a survival process, Accident Analysis and Prevention, 22:a sleeper berth, rest Accident Analysis and Prevention, 20:using survival theory, Accident Analysis and Prevention, 21:

  16. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    E-Print Network [OSTI]

    Hend Mohammed El Sayed Saad; Hesham Mohammed Mohammed Mansour; Moustafa Aziz Abd El Wahab

    2013-06-05T23:59:59.000Z

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and satisfactory agreement is found.

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

    SciTech Connect (OSTI)

    Not Available

    1988-12-15T23:59:59.000Z

    This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

  18. Superheated-steam test of ethylene propylene rubber cables using a simultaneous aging and accident environment

    SciTech Connect (OSTI)

    Bennett, P.R.; St. Clair, S.D.; Gilmore, T.W.

    1986-06-01T23:59:59.000Z

    The superheated-steam test exposed different ethylene propylene rubber (EPR) cables and insulation specimens to simultaneous aging and a 21-day simultaneous accident environment. In addition, some insulation specimens were exposed to five different aging conditions prior to the 21-day simultaneous accident simulation. The purpose of this superheated-steam test (a follow-on to the saturated-steam tests (NUREG/CR-3538)) was to: (1) examine electrical degradation of different configurations of EPR cables; (2) investigate differences between using superheated-steam or saturated-steam at the start of an accident simulation; (3) determine whether the aging technique used in the saturated-steam test induced artificial degradation; and (4) identify the constituents in EPR that affect moisture absorption.

  19. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    E-Print Network [OSTI]

    Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

    2013-01-01T23:59:59.000Z

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

  20. Assessment of CRBR core disruptive accident energetics

    SciTech Connect (OSTI)

    Theofanous, T.G.; Bell, C.R.

    1984-03-01T23:59:59.000Z

    The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly.

  1. A Review of Criticality Accidents 2000 Revision

    SciTech Connect (OSTI)

    Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

    2000-05-01T23:59:59.000Z

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

  2. WASTE-ACC: A computer model for analysis of waste management accidents

    SciTech Connect (OSTI)

    Nabelssi, B.K.; Folga, S.; Kohout, E.J.; Mueller, C.J.; Roglans-Ribas, J.

    1996-12-01T23:59:59.000Z

    In support of the U.S. Department of Energy`s (DOE`s) Waste Management Programmatic Environmental Impact Statement, Argonne National Laboratory has developed WASTE-ACC, a computational framework and integrated PC-based database system, to assess atmospheric releases from facility accidents. WASTE-ACC facilitates the many calculations for the accident analyses necessitated by the numerous combinations of waste types, waste management process technologies, facility locations, and site consolidation strategies in the waste management alternatives across the DOE complex. WASTE-ACC is a comprehensive tool that can effectively test future DOE waste management alternatives and assumptions. The computational framework can access several relational databases to calculate atmospheric releases. The databases contain throughput volumes, waste profiles, treatment process parameters, and accident data such as frequencies of initiators, conditional probabilities of subsequent events, and source term release parameters of the various waste forms under accident stresses. This report describes the computational framework and supporting databases used to conduct accident analyses and to develop source terms to assess potential health impacts that may affect on-site workers and off-site members of the public under various DOE waste management alternatives.

  3. Markov Model of Accident Progression at Fukushima Daiichi

    SciTech Connect (OSTI)

    Cuadra A.; Bari R.; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G.; Mubayi, V.; Pratt, T.; Yue, M.

    2012-11-11T23:59:59.000Z

    On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

  4. LESSONS LEARNED FROM A RECENT LASER ACCIDENT

    SciTech Connect (OSTI)

    Woods, Michael; /SLAC

    2011-01-26T23:59:59.000Z

    A graduate student received a laser eye injury from a femtosecond Ti:sapphire laser beam while adjusting a polarizing beam splitter optic. The direct causes for the accident included failure to follow safe alignment practices and failure to wear the required laser eyewear protection. Underlying root causes included inadequate on-the-job training and supervision, inadequate adherence to requirements, and inadequate appreciation for dimly visible beams outside the range of 400-700nm. This paper describes how the accident occurred, discusses causes and lessons learned, and describes corrective actions being taken.

  5. MELCOR accident consequence code system (MACCS)

    SciTech Connect (OSTI)

    Alpert, D.J.; Chanin, D.I.; Helton, J.C.; Ostmeyer, R.M.; Ritchie, L.T.

    1985-01-01T23:59:59.000Z

    Currently, the usefulness of reactor accident consequence assessments for providing guidance for planning and decision making is limited by the poor definition of uncertainties in predicted results. The MELCOR Accident Consequence Code System has been structured to facilitate performing uncertainty and sensitivity analyses. MACCS incorporates improved modeling capabilities in the treatment of variable or long duration releases, deposition modeling, dosimetry, emergency response, radiological health effects, and economic effects. At this writing (March 1985), the new code system has been completed and is undergoing testing, de-bugging, etc. Release of the first version of the full MELCOR code system, with associated documentation, is anticipated for the Autumn of 1985.

  6. The temporal effect of traffic violations and accidents on accident occurrence

    E-Print Network [OSTI]

    McKemie, Martha Susan

    1979-01-01T23:59:59.000Z

    THE TEMPORAL EFFECT OF TRAFFIC VIOLATIONS AND ACCIDENTS ON ACCIDENT OCCURRENCE A Thesis by . 1artha Susan McKemie Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER... OF SCIENCE December 1979 Major Subject: Industrial Engineering THE TEMPORAL El'FECT OF TRAI'FIC VIOIATIONS AND ACCIDENTS ON XCCIDENT OCCURPEENCE A Thesis by Martha Susan McKemie Approved as to style and content by: / ~J' (Chairman of Commi tee...

  7. Running Boundary Condition

    E-Print Network [OSTI]

    Satoshi Ohya; Makoto Sakamoto; Motoi Tachibana

    2013-01-28T23:59:59.000Z

    In this paper we argue that boundary condition may run with energy scale. As an illustrative example, we consider one-dimensional quantum mechanics for a spinless particle that freely propagates in the bulk yet interacts only at the origin. In this setting we find the renormalization group flow of U(2) family of boundary conditions exactly. We show that the well-known scale-independent subfamily of boundary conditions are realized as fixed points. We also discuss the duality between two distinct boundary conditions from the renormalization group point of view. Generalizations to conformal mechanics and quantum graph are also discussed.

  8. accident localisation system: Topics by E-print Network

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

    to designing performance monitoring and safety metrics. 1 Nancy Leveson 2004-01-01 14 The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson,...

  9. accident survival time: Topics by E-print Network

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

    person(s) involved in IncidentAccident: 1) Name New Hampshire, University of 2 Does Daylight Savings Time Affect Traffic Accidents? Texas A&M University - TxSpace Summary: This...

  10. accident issledovanie raspredeleniya: Topics by E-print Network

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

    Phone Home Phone Work Phone Exact Location : Date of Accident : AM PM Date accident treatment provided? Yes No Where Was time lost from work? Yes No If yes, how long? Could this...

  11. accident soderzhanie korotkozhivushchikh: Topics by E-print Network

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

    Phone Home Phone Work Phone Exact Location : Date of Accident : AM PM Date accident treatment provided? Yes No Where Was time lost from work? Yes No If yes, how long? Could this...

  12. PNNL Results from 2009 Silene Criticality Accident Dosimeter Intercomparison Exercise

    SciTech Connect (OSTI)

    Hill, Robin L.; Conrady, Matthew M.

    2010-06-30T23:59:59.000Z

    This document reports the results of testing of the Hanford Personnel Nuclear Accident Dosimeter (PNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on October 13, 14, and 15, 2009.

  13. Type B Accident Investigation, Response to the 24 Command Wildland...

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

    Type B Accident Investigation, Response to the 24 Command Wildland Fire on the Hanford Site, June 27-July 1, 2000 Type B Accident Investigation, Response to the 24 Command Wildland...

  14. accidents epidemiology trends: Topics by E-print Network

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

    8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Accident epidemiology and the US chemical industry: accident history and worst-case data from...

  15. accident du travail: Topics by E-print Network

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

    18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Bordereau de transmission accident du travail Mathematics Websites Summary: Bordereau de transmission accident du...

  16. Technical evaluation: 300 Area steam line valve accident

    SciTech Connect (OSTI)

    Not Available

    1993-08-01T23:59:59.000Z

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ``blanked off`` with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed.

  17. ACCIDENT ANALYSIS AND HAZARD ANALYSIS FOR HUMAN AND ORGANIZATIONAL FACTORS

    E-Print Network [OSTI]

    Leveson, Nancy

    culpable. An accident analysis method is needed that will guide the work, aid in the analysis of the role

  18. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect (OSTI)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01T23:59:59.000Z

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  19. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01T23:59:59.000Z

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  20. Impact of rainstorm and runoff modeling on predicted consequences of atmospheric releases from nuclear reactor accidents

    SciTech Connect (OSTI)

    Ritchie, L.T.; Brown, W.D.; Wayland, J.R.

    1980-05-01T23:59:59.000Z

    A general temperate latitude cyclonic rainstorm model is presented which describes the effects of washout and runoff on consequences of atmospheric releases of radioactive material from potential nuclear reactor accidents. The model treats the temporal and spatial variability of precipitation processes. Predicted air and ground concentrations of radioactive material and resultant health consequences for the new model are compared to those of the original WASH-1400 model under invariant meteorological conditions and for realistic weather events using observed meteorological sequences. For a specific accident under a particular set of meteorological conditions, the new model can give significantly different results from those predicted by the WASH-1400 model, but the aggregate consequences produced for a large number of meteorological conditions are similar.

  1. Technical basis document for the steam intrusion from interfacing systems accident

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-03-21T23:59:59.000Z

    This technical basis document was developed to support the Documented Safety Analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for the steam intrusion from interfacing systems representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

  2. Characterization of a nuclear accident dosimeter

    E-Print Network [OSTI]

    Burrows, Ronald Allen

    1995-01-01T23:59:59.000Z

    The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12-16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia...

  3. INTERNATIONAL STUDENT & SCHOLAR Accident & Sickness Insurance Plan

    E-Print Network [OSTI]

    Bordenstein, Seth

    and scholars participating in international educational programs outside of the United States. It is strongly an accident and sickness insurance plan for international students and scholars studying in the United States. The International Student & Scholar plan has a low monthly rate of $70 per person. WE'VE GOT YOU COVERED

  4. ANS severe accident program overview & planning document

    SciTech Connect (OSTI)

    Taleyarkhan, R.P.

    1995-09-01T23:59:59.000Z

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  5. L'accident la centrale nuclaire de Quelques explications scientifiques

    E-Print Network [OSTI]

    Skorobogatiy, Maksim

    L'accident à la centrale nucléaire de Fukushima Quelques explications scientifiques G. Marleau, J´eal, 18 mars 2011 L'accident `a la centrale nucl´eaire de Fukushima ­ 1/29 Accident de Fukushima 1 Contenu de Fukushima. 3. La puissance résiduelle. 4. Perte de refroidissement et conséquences. 5

  6. Policy 3240 Accident Review Committee 1 OLD DOMINION UNIVERSITY

    E-Print Network [OSTI]

    Policy 3240 ­ Accident Review Committee 1 OLD DOMINION UNIVERSITY University Policy Policy #3240 ACCIDENT REVIEW COMMITTEE Responsible Oversight Executive: Vice President for Administration and Finance vehicles for which ODU is responsible and the University's Accident Review Committee in the review

  7. HEALTH AND ACCIDENT INSURANCE VERIFICATION ******************** TO BE COMPLETED BY STUDENT ********************

    E-Print Network [OSTI]

    Jawitz, James W.

    HEALTH AND ACCIDENT INSURANCE VERIFICATION ******************** TO BE COMPLETED BY STUDENT Services Office of the university of Florida requires that s/he has health and accident insurance with your participating in study abroad activate hold health and accident insurance with a minimum coverage of $200

  8. For the mathematically accident prone student W Stephen Wilson

    E-Print Network [OSTI]

    Wilson, W. Stephen

    For the mathematically accident prone student by W Stephen Wilson Many students make the claim answers, whatever the reason for the incorrect answer. Students who are accident prone in mathematics. This is generally good advice for anyone, not just the accident prone. As problems get more and more complicated

  9. A New Accident Model for Engineering Safer Systems Nancy Leveson

    E-Print Network [OSTI]

    Leveson, Nancy

    A New Accident Model for Engineering Safer Systems Nancy Leveson Aeronautics and Astronautics Dept changes in the etiology of accidents and is creating a need for changes in the explanatory mechanisms used. We need better and less subjective understanding of why accidents occur and how to prevent future

  10. Structure Evolution of Dynamic Bayesian Network for Traffic Accident Detection

    E-Print Network [OSTI]

    Cho, Sung-Bae

    Structure Evolution of Dynamic Bayesian Network for Traffic Accident Detection Ju-Won Hwang, Young and the accuracy in a domain of the traffic accident detection. Keywords-structure of dynamic Bayesian network; Bayesian network, evolution I. INTRODUCTION Every year, traffic congestion and traffic accidents have been

  11. Annexes 195 13.11 Fecal Accident Plan

    E-Print Network [OSTI]

    Annexes 195 13.11 Fecal Accident Plan Residual and Contact Time Table Loose Stool Chlorine Residual and Contact Time Table Formed Stool Chlorine Residual mg/l or PPM Time Minutes 2 25 Sample Fecal Accident/spa at three locations to ensure proper mixing. Record fecal accidents in maintenance logs. Follow normal pool

  12. A STAMP ANALYSIS OF THE LEX COMAIR 5191 ACCIDENT

    E-Print Network [OSTI]

    Leveson, Nancy

    A STAMP ANALYSIS OF THE LEX COMAIR 5191 ACCIDENT Thesis submitted in partial fulfilment;A STAMP ANALYSIS OF THE LEX COMAIR 5191 ACCIDENT Paul S. Nelson 2 #12;Acknowledgements I want pressure" (Dekker, 2007, p. 131) A new, holistic systems perspective, accident model is used for analysis

  13. COLUMBIA UNIVERSITY Departmental Accident Report Form for Worker's Compensation Benefits

    E-Print Network [OSTI]

    Jia, Songtao

    COLUMBIA UNIVERSITY Departmental Accident Report Form for Worker's Compensation Benefits EMPLOYEE___________ ACCIDENT DATA (to be completed by employee) Date of Injury_____/_____/____ Time of Injury the employee How did the injury or illness occur? (Describe fully the events that caused the accident) Describe

  14. DEVELOPMENT AND USE OF A DIRECTORY OF ACCIDENT DATABASES INVOLVING

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    DEVELOPMENT AND USE OF A DIRECTORY OF ACCIDENT DATABASES INVOLVING CHEMICALS J.P.Pineau Institut from end-users of accident data who need validated data for dealing with risk assessment in which Data collection Data analysis, Reliability, Uncertainty, Accident, Hazardous material, Risk analysis

  15. CLAIMANT AUTO ACCIDENT REPORT For Completion by Driver

    E-Print Network [OSTI]

    Tullos, Desiree

    CLAIMANT AUTO ACCIDENT REPORT For Completion by Driver D E P A R T M E N T O F A D M I N I S T R Address City State Zip For what purpose was car being used at time of accident? Has damage been repaired signals did you give? Other Driver? Who investigated? Who Cited and Why? Describe Accident CONTINUE

  16. Scar sarcoidosis with a 50-year interval between an accident and onset of lesions

    E-Print Network [OSTI]

    Jr, Hiram Larangeira de Almeida; Fiss, Roberto Coswig

    2008-01-01T23:59:59.000Z

    year interval between an accident and onset of lesions Hiramreported in scars of accidents [ 2 ], herpes zoster [ 1 ],

  17. Risk assessment of severe accident-induced steam generator tube rupture

    SciTech Connect (OSTI)

    NONE

    1998-03-01T23:59:59.000Z

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC`s Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs.

  18. World Views From fragmentation

    E-Print Network [OSTI]

    .......................................................11 2. The Seven Components of a World View...................................................... 20 3. The Unity of the Seven Sub........................................... 25 5. The Purpose of the group `Worldviews

  19. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    SciTech Connect (OSTI)

    Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

    2011-06-21T23:59:59.000Z

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

  20. REAC/TS Radiation Accident Registry: An Overview

    SciTech Connect (OSTI)

    Doran M. Christensen, DO, REAC /TS Associate Director and Staff Physician Becky Murdock, REAC/TS Registry and Health Physics Technician

    2012-12-12T23:59:59.000Z

    Over the past four years, REAC/TS has presented a number of case reports from its Radiation Accident Registry. Victims of radiological or nuclear incidents must meet certain dose criteria for an incident to be categorized as an “accident” and be included in the registry. Although the greatest numbers of “accidents” in the United States that have been entered into the registry involve radiation devices, the greater percentage of serious accidents have involved sealed sources of one kind or another. But if one looks at the kinds of accident scenarios that have resulted in extreme consequence, i.e., death, the greater share of deaths has occurred in medical settings.

  1. NEWS & VIEWS Glass dynamics

    E-Print Network [OSTI]

    Weeks, Eric R.

    NEWS & VIEWS Glass dynamics Diverging views on glass transition Gregory B. mc.mckenna@ttu.edu T he glass transition is one of the most intriguing phenomena in the world of soft condensed matter. Despite decades of study, many aspects of the behaviour of glass-forming liquids remain elusive

  2. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01T23:59:59.000Z

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilitiesmore »while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.« less

  3. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01T23:59:59.000Z

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilities while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.

  4. Rflexions sur le transfert mthodologique de l'analyse qualitative d'accidents de la circulation routire issue de l'tude dtaille des accidents (EDA) franaise aux

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    transfer for qualitative road accident analysis obtained from French Detailed Accident Studies (DAS the comprehensive accident analysis methodologies used in developed countries provide an understanding of the origin accident studies (DASs) and their adaptation to the analysis of accident reports. Colombia has

  5. The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson, Ph.D.; Massachusetts Institute of Technology; Cambridge, Massachusetts

    E-Print Network [OSTI]

    Leveson, Nancy

    The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson, Ph.D.; University of Victoria; Victoria, Canada Keywords: accident analysis, accident models Abstract In another paper presented at this conference, Leveson describes a new accident model based on systems theory [2

  6. HOW TO REPORT AN ACCIDENT, INCIDENT OR NEAR MISS 1. Notify your supervisor or lab manager as soon as possible of your accident, incident, or

    E-Print Network [OSTI]

    Borenstein, Elhanan

    HOW TO REPORT AN ACCIDENT, INCIDENT OR NEAR MISS 1. Notify your supervisor or lab manager as soon as possible of your accident, incident, or near miss. 2. Fill out the online accident report (OARS) form://www.ehs.washington.edu/ohsoars/index.shtm. The supervisor, lab manager, or person who had the accident can fill out the form. 3. For any serious accidents

  7. US Department of Energy Chernobyl accident bibliography

    SciTech Connect (OSTI)

    Kennedy, R A; Mahaffey, J A; Carr, F Jr

    1992-04-01T23:59:59.000Z

    This bibliography has been prepared by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) Office of Health and Environmental Research to provide bibliographic information in a usable format for research studies relating to the Chernobyl nuclear accident that occurred in the Ukrainian Republic, USSR in 1986. This report is a product of the Chernobyl Database Management project. The purpose of this project is to produce and maintain an information system that is the official United States repository for information related to the accident. Two related products prepared for this project are the Chernobyl Bibliographic Search System (ChernoLit{trademark}) and the Chernobyl Radiological Measurements Information System (ChernoDat). This report supersedes the original release of Chernobyl Bibliography (Carr and Mahaffey, 1989). The original report included about 2200 references. Over 4500 references and an index of authors and editors are included in this report.

  8. Understanding the Columbia Space Shuttle Accident

    SciTech Connect (OSTI)

    Osheroff, Doug (Stanford University) [Stanford University

    2004-06-16T23:59:59.000Z

    On February 1, 2003, the NASA space shuttle Columbia broke apart during re-entry over East Texas at an altitude of 200,000 feet and a velocity of approximately 12,000 mph. All aboard perished. Prof. Osheroff was a member of the board that investigated the origins of this accident, both physical and organizational. In his talk he will describe how the board was able to determine with almost absolute certainty the physical cause of the accident. In addition, Prof. Osherhoff will discuss its organizational and cultural causes, which are rooted deep in the culture of the human spaceflight program. Why did NASA continue to fly the shuttle system despite the persistent failure of a vital sub-system that it should have known did indeed pose a safety risk on every flight? Finally, Prof. Osherhoff will touch on the future role humans are likely to play in the exploration of space.

  9. Coupled thermal analysis applied to the study of the rod ejection accident

    SciTech Connect (OSTI)

    Gonnet, M. [AREVA NP, TOUR AREVA - 1 Place Jean MILLIER, 92084 Paris La Defense Cedex (France)

    2012-07-01T23:59:59.000Z

    An advanced methodology for the assessment of fuel-rod thermal margins under RIA conditions has been developed by AREVA NP SAS. With the emergence of RIA analytical criteria, the study of the Rod Ejection Accident (REA) would normally require the analysis of each fuel rod, slice by slice, over the whole core. Up to now the strategy used to overcome this difficulty has been to perform separate analyses of sampled fuel pins with conservative hypotheses for thermal properties and boundary conditions. In the advanced methodology, the evaluation model for the Rod Ejection Accident (REA) integrates the node average fuel and coolant properties calculation for neutron feedback purpose as well as the peak fuel and coolant time-dependent properties for criteria checking. The calculation grid for peak fuel and coolant properties can be specified from the assembly pitch down to the cell pitch. The comparative analysis of methodologies shows that coupled methodology allows reducing excessive conservatism of the uncoupled approach. (authors)

  10. Risk Estimation Methodology for Launch Accidents.

    SciTech Connect (OSTI)

    Clayton, Daniel James; Lipinski, Ronald J.; Bechtel, Ryan D.

    2014-02-01T23:59:59.000Z

    As compact and light weight power sources with reliable, long lives, Radioisotope Power Systems (RPSs) have made space missions to explore the solar system possible. Due to the hazardous material that can be released during a launch accident, the potential health risk of an accident must be quantified, so that appropriate launch approval decisions can be made. One part of the risk estimation involves modeling the response of the RPS to potential accident environments. Due to the complexity of modeling the full RPS response deterministically on dynamic variables, the evaluation is performed in a stochastic manner with a Monte Carlo simulation. The potential consequences can be determined by modeling the transport of the hazardous material in the environment and in human biological pathways. The consequence analysis results are summed and weighted by appropriate likelihood values to give a collection of probabilistic results for the estimation of the potential health risk. This information is used to guide RPS designs, spacecraft designs, mission architecture, or launch procedures to potentially reduce the risk, as well as to inform decision makers of the potential health risks resulting from the use of RPSs for space missions.

  11. Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

    SciTech Connect (OSTI)

    Tusheva, P.; Schaefer, F.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden (Germany)

    2012-07-01T23:59:59.000Z

    The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

  12. POWER LEVEL EFFECT IN A PWR ROD EJECTION ACCIDENT.

    SciTech Connect (OSTI)

    DIAMOND,D.J.; BROMLEY,B.P.; ARONSON,A.L.

    2002-10-07T23:59:59.000Z

    The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS, a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation.

  13. NEWS AND VIEWS PERSPECTIVE

    E-Print Network [OSTI]

    Mahler, D. Luke

    NEWS AND VIEWS PERSPECTIVE Niche diversification follows key innovation in Antarctic fish radiation Oxford Street, Cambridge MA 02138, USA Antarctic notothenioid fishes provide a fascinating evolu- tionary diversification has occurred repeatedly and in parallel. Keywords: community ecology, fish, macroevolution, phylo

  14. Forward viewing OCT endomicroscopy

    E-Print Network [OSTI]

    Liang, Kaicheng

    2014-01-01T23:59:59.000Z

    A forward viewing fiber optic-based imaging probe device was designed and constructed for use with ultrahigh speed optical coherence tomography in the human gastrointestinal tract. The light source was a MEMS-VCSEL at 1300 ...

  15. A framework for the assessment of severe accident management strategies

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

    1993-09-01T23:59:59.000Z

    Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

  16. Interdisciplinary Institute for Innovation Le risque d'accident nuclaire

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Interdisciplinary Institute for Innovation Le risque d'accident nucléaire majeur : calcul et-27Feb2013 #12;Le risque d'accident nucléaire majeur : calcul et perception des probabilités1 François Lévêque L'accident de Fukushima Daiichi s'est produit le 11 mars 2011. Cette catastrophe nucléaire

  17. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect (OSTI)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01T23:59:59.000Z

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

  18. assigned accident investigation: Topics by E-print Network

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

    33 Long-term investigations of radiocaesium activity concentrations in carps in north Croatia after the Chernobyl accident CERN Preprints Summary: Long-term investigations of...

  19. accident source term: Topics by E-print Network

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

    42 Long-term investigations of radiocaesium activity concentrations in carps in north Croatia after the Chernobyl accident CERN Preprints Summary: Long-term investigations of...

  20. accident investigation: Topics by E-print Network

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

    26 Long-term investigations of radiocaesium activity concentrations in carps in north Croatia after the Chernobyl accident CERN Preprints Summary: Long-term investigations of...

  1. accident source terms: Topics by E-print Network

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

    42 Long-term investigations of radiocaesium activity concentrations in carps in north Croatia after the Chernobyl accident CERN Preprints Summary: Long-term investigations of...

  2. Type A Accident Investigation of the March 16, 2000, Plutonium...

    Office of Environmental Management (EM)

    Multiple Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico Type A Accident Investigation of the March 16, 2000, Plutonium-238 Multiple Intake...

  3. Type B Accident Investigation Report on the Exertional Heat Illnesses...

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

    Heat Illnesses during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July 13, 2006 Type B Accident Investigation Report on the Exertional Heat...

  4. Dose estimates in a loss of lead shielding truck accident.

    SciTech Connect (OSTI)

    Dennis, Matthew L.; Osborn, Douglas M.; Weiner, Ruth F.; Heames, Terence John (Alion Science & Technology Albuquerque, NM)

    2009-08-01T23:59:59.000Z

    The radiological transportation risk & consequence program, RADTRAN, has recently added an updated loss of lead shielding (LOS) model to it most recent version, RADTRAN 6.0. The LOS model was used to determine dose estimates to first-responders during a spent nuclear fuel transportation accident. Results varied according to the following: type of accident scenario, percent of lead slump, distance to shipment, and time spent in the area. This document presents a method of creating dose estimates for first-responders using RADTRAN with potential accident scenarios. This may be of particular interest in the event of high speed accidents or fires involving cask punctures.

  5. affecting reactor accident: Topics by E-print Network

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

    METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION 2 Does Daylight Savings Time Affect Traffic Accidents? Texas A&M University - TxSpace Summary: This...

  6. Type A Accident Investigation Board Report on the February 20...

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

    February 20, 1996, Fall Fatality at the Radioactive Waste Management Complex Transuranic Storage Area - Retrieval Enclosure, Idaho National Engineering Laboratory Type A Accident...

  7. Microsoft Word - Case Study for Enhanced Accident Tolerance Design...

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

    2355 Case Study for Enhanced Accident Tolerance Design Changes Steven Prescott Curtis Smith Tony Koonce June 2014 DISCLAIMER This information was prepared as an account of work...

  8. Type B Accident Investigation Board Report on the October 15...

    Energy Savers [EERE]

    on the October 15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental Laboratory Type B Accident Investigation Board...

  9. accident loca testing: Topics by E-print Network

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

    for the degree ol' MASTER OF SCIENCE May 1992 Major Subject: Nuclear Engineering SIMULATION OF A SMALL BREAK LOSS OF COOLANT ACCIDENT CONDUCTED AT THE BETHSY INTEGRAL TEST...

  10. Sec. Herrington Leads Delegation in Response to Chernobyl Accident...

    National Nuclear Security Administration (NNSA)

    Sec. Herrington Leads Delegation in Response to Chernobyl Accident | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  11. Type B Accident Investigation Report of the October 28, 2004...

    Energy Savers [EERE]

    of the October 28, 2004, Burn Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA Type B Accident Investigation Report...

  12. Type B Accident Investigation Board Report of the September 29...

    Energy Savers [EERE]

    at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna, New, York Type B Accident Investigation Board Report of the September 29, 2010,...

  13. accident prevention manual: Topics by E-print Network

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

    Wu, Mingshen 9 Chest--Manual Defrost Models Biology and Medicine Websites Summary: old refrigerator or freezer, please follow the instructions below to help prevent accidents....

  14. accident management programme: Topics by E-print Network

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

    ACCIDENT FIRE POLLUTION "NEAR MISS immediately after the occurrence. 3 Material damage or pollution Total volume of mercury spillage was approximately 200 ml. Of that volume,...

  15. accident management programmes: Topics by E-print Network

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

    ACCIDENT FIRE POLLUTION "NEAR MISS immediately after the occurrence. 3 Material damage or pollution Total volume of mercury spillage was approximately 200 ml. Of that volume,...

  16. Type B Accident Investigation Board Report, May 8, 2004, Exothermic...

    Energy Savers [EERE]

    Report, May 8, 2004, Exothermic Metal Reactor Event During Sodium Transfer Activities, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board...

  17. Accident Investigation of the September 20, 2012 Fatal Fall from...

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

    September 20, 2012 Fatal Fall from the Dworshak-Taft 1 Transmission Tower, at the Bonneville Power Marketing Administration Accident Investigation of the September 20, 2012 Fatal...

  18. Accident Investigation of the July 30, 2013, Electrical Fatality...

    Energy Savers [EERE]

    to the Secretary of Labor Accident Investigation of the September 20, 2012 Fatal Fall from the Dworshak-Taft 1 Transmission Tower, at the Bonneville Power Marketing Administration...

  19. Type B Accident Investigation At Washington Closure Hanford,...

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

    Fall Injury on July 1, 2009, At The 336 Building, Hanford Site, Washington Type B Accident Investigation At Washington Closure Hanford, LLC, Employee Fall Injury on July 1,...

  20. Type B Accident Investigation of the Arc Flash at Brookhaven...

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

    Arc Flash at Brookhaven National Laboratory, April 14, 2006 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 February 10, 2006 An...

  1. accidents involving external: Topics by E-print Network

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

    25 Next Page Last Page Topic Index 1 Development and use of the ESReDA directory of accident databases involving chemicals Computer Technologies and Information Sciences Websites...

  2. Type B Accident Investigation Board Report of the Brookhaven...

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

    National Laboratory Employee Injury at Building 1005H on October 9, 2009 Type B Accident Investigation Board Report of the Brookhaven National Laboratory Employee Injury at...

  3. accident resistant container: Topics by E-print Network

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

    failure of thermal barrier coatings (TBCs) driven by thickening Wadley, Haydn 2 OTHER ACCIDENT?24. ANY PERSON WHO KNOWINGLY AND WITH INTENT TO DEFRAUD ANY INSURANCE COMPANY OR...

  4. accident diagrams: Topics by E-print Network

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

    study of groups (see 22 or 26) Victor Guba; Mark Sapir 1996-01-01 2 AUTOMOBILE ACCIDENT REPORT Department of Financial Services Geosciences Websites Summary: . 0103 (USE...

  5. Accident Investigation of the December 11, 2013, Integrated Device...

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

    Accidental Discharge at the Sandia National Laboratory Site 9920, Albuquerque, NM Accident Investigation of the December 11, 2013, Integrated Device Fireset and Detonator...

  6. Type B Accident Investigation of the August 22, 2000, Injury...

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

    Type B Accident Investigation of the August 22, 2000, Injury Resulting From Violent Exothermic Chemical Reaction at the Portsmouth Gaseous Diffusion Plant, X-701B Site Type B...

  7. accident response calculations: Topics by E-print Network

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

    25 Next Page Last Page Topic Index 1 Primary Responsibilities 1. Identify potential accident hazards. Materials Science Websites Summary: Primary Responsibilities 1. Identify...

  8. Type B Accident Investigation Board Report of the Bechtel Jacobs...

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

    at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board Report of the Bechtel Jacobs Company, LLC Employee Fall Injury on...

  9. Severe Accident Studies | Department of Energy

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transfer toSensor Technologies for a SmartSevere Accident Studies

  10. Ramkrishna Mukherjee. Uganda: An Historical Accident?: Class, Natona, State Formation. Trenton, New Jersey: Africa World Press, 1985 281pp.

    E-Print Network [OSTI]

    Isabirye, Stephen B.

    1989-01-01T23:59:59.000Z

    Trenton, Historical Accident? : Class, Natona, New Jersey:in Mukherjee Historical Accident. analyzes the "poUticalare not an "historical accident." War, Violence and Children

  11. Uncertainty quantification for accident management using ACE surrogates

    SciTech Connect (OSTI)

    Varuttamaseni, A.; Lee, J. C. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48109 (United States); Youngblood, R. W. [Idaho National Laboratory, Idaho Falls, ID 83415-3870 (United States)

    2012-07-01T23:59:59.000Z

    The alternating conditional expectation (ACE) regression method is used to generate RELAP5 surrogates which are then used to determine the distribution of the peak clad temperature (PCT) during the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed (F and B) operation in the Zion-1 nuclear power plant. The construction of the surrogates assumes conditional independence relations among key reactor parameters. The choice of parameters to model is based on the macroscopic balance statements governing the behavior of the reactor. The peak clad temperature is calculated based on the independent variables that are known to be important in determining the success of the F and B operation. The relationship between these independent variables and the plant parameters such as coolant pressure and temperature is represented by surrogates that are constructed based on 45 RELAP5 cases. The time-dependent PCT for different values of F and B parameters is calculated by sampling the independent variables from their probability distributions and propagating the information through two layers of surrogates. The results of our analysis show that the ACE surrogates are able to satisfactorily reproduce the behavior of the plant parameters even though a quasi-static assumption is primarily used in their construction. The PCT is found to be lower in cases where the F and B operation is initiated, compared to the case without F and B, regardless of the F and B parameters used. (authors)

  12. Material Selection for Accident Tolerant Fuel Cladding

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    none,

    2014-07-01T23:59:59.000Z

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200°C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steammore »and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ?’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated. Keywords: Accident tolerant LWR Fuel cladding, FeCrAl, Mo, Ti2AlC, Al2O3, high temperature steam oxidation resistance« less

  13. Material Selection for Accident Tolerant Fuel Cladding

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    none,

    2014-07-01T23:59:59.000Z

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200°C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ?’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated. Keywords: Accident tolerant LWR Fuel cladding, FeCrAl, Mo, Ti2AlC, Al2O3, high temperature steam oxidation resistance

  14. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    SciTech Connect (OSTI)

    Cahalan, J.; Wigeland, R. (Argonne National Lab., IL (USA)); Friedel, G. (Internationale Atomreaktorbau GmbH (INTERATOM), Bergisch Gladbach (Germany, F.R.)); Kussmaul, G.; Royl, P. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.)); Moreau, J. (CEA Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France)); Perks, M. (UKAEA Risley Nuclear Power Development Establishment (UK)

    1990-01-01T23:59:59.000Z

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs.

  15. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01T23:59:59.000Z

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  16. Engineering Aerial view of

    E-Print Network [OSTI]

    Yang, Junfeng

    -neutral Torus 2 Climate Change 4 Combustion and Catalysis Laboratory #12;4 5 1Engineering Revolution 5 #12;6 7Columbia Engineering Plus #12;1 1 2 3 4 5 6 Aerial view of Columbia campus with Columbia Engineering-a liated buildings highlighted in blue Columbia Engineering Plus Engineering Revolution 4

  17. Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video

    Broader source: Energy.gov [DOE]

    This course that provides an overview of the fundamentals of accident investigation. The course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE O 225.1B, Accident Investigations.

  18. Using a town’s GIS project to create a deer-vehicle accident management plan

    E-Print Network [OSTI]

    Rogers, Elizabeth I.

    2003-01-01T23:59:59.000Z

    TO CREATE A DEER-VEHICLE ACCIDENT MANAGEMENT PLAN Elizabethhigh numbers of deer-vehicle accidents (DVAs) on a landscapeto provide an assessment of accident risk in time and space.

  19. Road traffic accidents in Kathmandu¿an hour of education yields a glimmer of hope

    E-Print Network [OSTI]

    Basnet, Bibhusan; Vohra, Rais; Bhandari, Amit; Pandey, Subash

    2013-01-01T23:59:59.000Z

    et al. : Road traffic accidents in Kathmandu— an hour ofOpen Access Road traffic accidents in Kathmandu—an hour ofnumber of road traffic accidents in the year 2012 decreased

  20. accident victims bio-indicateurs: Topics by E-print Network

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

    Every year, traffic congestion and traffic accidents have been Cho, Sung-Bae 119 The Analysis of a Friendly Fire Accident using a Systems Model of Accidents* N.G. Leveson,...

  1. Estimating Pedestrian Accident Exposure: Approaches to a Statewide Pedestrian Exposure Database

    E-Print Network [OSTI]

    Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01T23:59:59.000Z

    Pedestrian Exposure to Risk of Road Accident in New Zealand.Accident Analysis and Prevention, Vol. 27, No. 3, 1995, pp.Automated Traffic Accident Surveillance and Analysis System,

  2. UNIVERSITY OF TORONTO ACCIDENT/INCIDENT/OCCUPATIONAL DISEASE REPORT FOR EMPLOYEES

    E-Print Network [OSTI]

    Kronzucker, Herbert J.

    UNIVERSITY OF TORONTO ACCIDENT/INCIDENT/OCCUPATIONAL DISEASE REPORT FOR EMPLOYEES RELEVANT SECTIONS: _______________________________________ NAME OF SUPERVISOR TO WHOM ACCIDENT WAS REPORTED: _________________________________ TELEPHONE: _____________________ IF THERE WAS A DELAY IN REPORTING THIS ACCIDENT, LIST REASON

  3. STUDENT / VISITOR ACCIDENT REPORT FORM nco/revised 10/06/03

    E-Print Network [OSTI]

    Azevedo, Ricardo

    STUDENT / VISITOR ACCIDENT REPORT FORM nco/revised 10/06/03 (To Be Completed By Individual Involved In Accident) 1. Name: ________________________________________ Student ID or DL No.: _______________________ 2 No - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 6. Date of Accident: ___________________ Day of Week: _______________________ Time: ____________ 7

  4. Technical basis for the aboveground structure failure and associated represented hazardous conditions

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-07-25T23:59:59.000Z

    This technical basis document describes the risk binning process and the technical basis for assigning risk bins for the aboveground structure failure representative accident and associated represented hazardous conditions. This document was developed to support the documented safety analysis.

  5. accident analysis codes: Topics by E-print Network

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

    accident analysis codes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Analysis of accidents during...

  6. Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)

    SciTech Connect (OSTI)

    POWERS, T.B.

    2000-03-20T23:59:59.000Z

    By using simple frequency calculations and fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The following are the design basis accidents: Mechanical damage of MCO; Gaseous release from the MCO; MCO internal hydrogen deflagration; MCO external hydrogen deflagration; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

  7. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    CROWE, R.D.

    1999-09-09T23:59:59.000Z

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

  8. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    1999-10-20T23:59:59.000Z

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  9. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

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

    2000-03-23T23:59:59.000Z

    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.

  10. accident analysis documentation: Topics by E-print Network

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

    accident analysis documentation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Analysis of accidents...

  11. Berkeley Lab Accident Statistics Through December 31, 2008

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through December 31, 2008 These slides are updated on a monthly Goal DART Goal 1.17 #12;8 LBNL vs DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3

  12. accident proneness: Topics by E-print Network

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

    accident proneness First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Accident proneness as an expression...

  13. accident analysis: Topics by E-print Network

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

    accident analysis First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Analysis of accidents during flashing...

  14. accident atuacao da: Topics by E-print Network

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

    accident atuacao da First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 DIRECTORY OF ESReDA ACCIDENT...

  15. accident lofa analysis: Topics by E-print Network

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

    accident lofa analysis First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 MELCOR ACCIDENT ANALYSIS FOR...

  16. Berkeley Lab Accident Statistics Through November 30, 2008

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through November 30, 2008 These slides are updated on a monthly Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1

  17. accident precursor analysis: Topics by E-print Network

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

    accident precursor analysis First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Analysis of accidents...

  18. aircraft accident victims: Topics by E-print Network

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

    aircraft accident victims First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Cyclistes Victimes d'Accident...

  19. accident prone locations: Topics by E-print Network

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

    accident prone locations First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Accident proneness as an...

  20. Berkeley Lab Accident Statistics Through November 30, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through November 30, 2009 These slides are updated on a monthly Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1

  1. accident severity: Topics by E-print Network

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

    accident severity First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Accidents on the campus Severe...

  2. accident proneness prospect: Topics by E-print Network

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

    accident proneness prospect First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Accident proneness as an...

  3. accident consequences health: Topics by E-print Network

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

    accident consequences health First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 HEALTH AND ACCIDENT...

  4. Berkeley Lab Accident Statistics Through August 31, 2008

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through August 31, 2008 These slides are updated on a monthly 1.17 #12;7 LBNL vs DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3

  5. Berkeley Lab Accident Statistics Through April 30, 2010

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through April 30, 2010 These slides are updated on a monthly Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1

  6. accident characteristics: Topics by E-print Network

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

    accident characteristics First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 guia accidents BSICA ...

  7. alternative accident sequences: Topics by E-print Network

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

    alternative accident sequences First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Beyond Normal Accidents...

  8. Berkeley Lab Accident Statistics Through May 31, 2010

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through May 31, 2010 These slides are updated on a monthly basis DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2

  9. Berkeley Lab Accident Statistics Through June 30, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through June 30, 2009 These slides are updated on a monthly Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1

  10. Berkeley Lab Accident Statistics Through January 31, 2010

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through January 31, 2010 These slides are updated on a monthly Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28

  11. accident types: Topics by E-print Network

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

    accident types First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Risk Advisor for Car Accidents Javier...

  12. Berkeley Lab Accident Statistics Through October 31, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through October 31, 2009 These slides are updated on a monthly;8 LBNL vs DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2

  13. accident compensation insurance: Topics by E-print Network

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

    accident compensation insurance First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Group Accident...

  14. Berkeley Lab Accident Statistics Through September 30, 2008

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through September 30, 2008 These slides are updated on a monthly.17 #12;7 LBNL vs DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3

  15. accident situation study: Topics by E-print Network

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

    accident situation study First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Hypothetical Reactor Accident...

  16. accident exposure: Topics by E-print Network

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

    accident exposure First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Estimating Pedestrian Accident...

  17. Berkeley Lab Accident Statistics Through April 30, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through April 30, 2009 These slides are updated on a monthly Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2. 93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1

  18. accident sequence analyses: Topics by E-print Network

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

    accident sequence analyses First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Analysing Aviation Accidents...

  19. accident insurance: Topics by E-print Network

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

    accident insurance First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Group Accident Insurance Certificate...

  20. Berkeley Lab Accident Statistics Through December 31, 2010

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through December 31, 2010 These slides are updated on a monthly.17 #12;9 LBNL vs DOE Contractor Rates Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3

  1. aircraft accident cases: Topics by E-print Network

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

    accident cases First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 HOW PAST LOSS OF CONTROL ACCIDENTS MAY...

  2. accident dosimetry systems: Topics by E-print Network

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

    accident dosimetry systems First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 A New Accident Model for...

  3. accident loca based: Topics by E-print Network

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

    accident loca based First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 A GIS based traffic accident data...

  4. Statistical evaluation of design-error related accidents

    SciTech Connect (OSTI)

    Ott, K.O.; Marchaterre, J.F.

    1980-01-01T23:59:59.000Z

    In a recently published paper (Campbell and Ott, 1979), a general methodology was proposed for the statistical evaluation of design-error related accidents. The evaluation aims at an estimate of the combined residual frequency of yet unknown types of accidents lurking in a certain technological system. Here, the original methodology is extended, as to apply to a variety of systems that evolves during the development of large-scale technologies. A special categorization of incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of the nuclear power reactor technology, considering serious accidents that involve in the accident-progression a particular design inadequacy.

  5. A Technique for Showing Causal Arguments in Accident Reports C. W. Johnson; University of Glasgow; Glasgow, Scotland, UK

    E-Print Network [OSTI]

    Johnson, Chris

    A Technique for Showing Causal Arguments in Accident Reports C. W. Johnson; University of Glasgow: causes, accidents, logic, argument, visualization, road traffic accidents Abstract In the prototypical accident report, specific findings, particularly those related to causes and contributing factors

  6. MELCOR accident analysis for ARIES-ACT

    SciTech Connect (OSTI)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01T23:59:59.000Z

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  7. Angular dependence of a simple accident dosimeter

    SciTech Connect (OSTI)

    Devine, R. T. (Robert T.); Romero, L. L. (Leonard L.); Olsher, R. H. (Richard H.)

    2004-01-01T23:59:59.000Z

    A simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. The threshold cross sections for the detectors on a phantom were calculated. The resulting doses assigned agree well with exposures made to three critical assemblies. In this study the angular dependence on a phantom is studied and compared with measurements taken on the GODIVA reactor. The dosimeter positions on the phantom are facing the source, on the back and the side. In previous papers the modeling of a simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. The conclusion made was that most of the neutron dose from criticality assemblies results from the high energy neutron fluences determined by the sulfur and indium detectors. The results using doses measured from the GODIVA, SHEBA, and bare and lead shielded SILENE reactors confirmed this. The angular dependence of an accident dosemeter is of interest in evaluating the exposure of personnel. To investigate this effect accident dosemeters were placed on a phantom and exposed to the GODIVA reactor at phantom orientations of 0{sup o}, 45{sup o}, 90{sup o}, 135{sup o}, and 180{sup o} to the assembly center line.

  8. Stereoscopic optical viewing system

    DOE Patents [OSTI]

    Tallman, Clifford S. (Walnut Creek, CA)

    1987-01-01T23:59:59.000Z

    An improved optical system which provides the operator a stereoscopic viewing field and depth of vision, particularly suitable for use in various machines such as electron or laser beam welding and drilling machines. The system features two separate but independently controlled optical viewing assemblies from the eyepiece to a spot directly above the working surface. Each optical assembly comprises a combination of eye pieces, turning prisms, telephoto lenses for providing magnification, achromatic imaging relay lenses and final stage pentagonal turning prisms. Adjustment for variations in distance from the turning prisms to the workpiece, necessitated by varying part sizes and configurations and by the operator's visual accuity, is provided separately for each optical assembly by means of separate manual controls at the operator console or within easy reach of the operator.

  9. Stereoscopic optical viewing system

    DOE Patents [OSTI]

    Tallman, C.S.

    1986-05-02T23:59:59.000Z

    An improved optical system which provides the operator with a stereoscopic viewing field and depth of vision, particularly suitable for use in various machines such as electron or laser beam welding and drilling machines. The system features two separate but independently controlled optical viewing assemblies from the eyepiece to a spot directly above the working surface. Each optical assembly comprises a combination of eye pieces, turning prisms, telephoto lenses for providing magnification, achromatic imaging relay lenses and final stage pentagonal turning prisms. Adjustment for variations in distance from the turning prisms to the workpiece, necessitated by varying part sizes and configurations and by the operator's visual accuity, is provided separately for each optical assembly by means of separate manual controls at the operator console or within easy reach of the operator.

  10. Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

    SciTech Connect (OSTI)

    Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)) [Sandia National Labs., Albuquerque, NM (USA); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)) [GRAM, Inc., Albuquerque, NM (USA); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA)) [Arizona State Univ., Tempe, AZ (USA)

    1990-12-01T23:59:59.000Z

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.

  11. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect (OSTI)

    Nelson, Andrew T. [Los Alamos National Laboratory

    2012-07-24T23:59:59.000Z

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  12. Evaluation of severe accident risk during mid-loop operation at Surry unit-1

    SciTech Connect (OSTI)

    Mubayi, V.; Jo, J.; Lin, C.C.; Neymotin, L.; Pratt, W.T.

    1996-06-01T23:59:59.000Z

    In the past most probabilistic risk assessments (PRAs) of severe accidents in nuclear power plants have considered initiating events which could potentially lead to core damage and containment failure during normal full power operation. However, recent studies and operational experience during periods while plants were shutdown for maintenance or refueling indicated that potential accidents initiated during low power operation or shutdown conditions could also potentially become important contributors to risk. In 1989, the Nuclear Regulatory Commission (NRC) began an extensive program to assess the risk during low power and shutdown operation. Two plants, Surry (a pressurized water reactor, PWR) and Grand Gulf (a boiling water reactor ,BWR) were selected as the plants to be studied.This paper describes an analysis of accident progression and offsite consequences (level 3 PRA) carried out for the Surry plant. The focus of the level 3 PRA was on mid-loop operation, which is a plant operational state (POS) that can occur while the plant is shutdown for maintenance or refueling. Mid-loop refers to a configuration when the reactor coolant system is lowered to the mid-plane of the hot leg to allow essential maintenance to be performed. This operational state was selected after an initial coarse screening study indicated that reduced inventory during mid-loop operation could pose higher risk than other POSs.

  13. Hazardous-material accidents near nuclear power plants: an evaluation of analyses and approaches

    SciTech Connect (OSTI)

    Kot, C.A.; Lin, H.C.; van Erp, J.B.; Eichler, T.V.; Wiedermann, A.H.

    1983-10-01T23:59:59.000Z

    The state of knowledge concerning postulated accidents involving offsite hazardous materials in the vicinity of nuclear power plants is critically evaluated. This effort is part of a study to analyze the potential effects of offsite hazards upon the safety of nuclear power plants and to develop a technical basis for the assessment of siting approaches. The evaluation includes consideration of data bases and statistics of hazardous materials and accidents involving them, deterministic aspects of possible material dispersion and threat environments, the susceptibility and vulnerability of vital plant systems, and a critical review of past licensing experience and regulatory practice with respect to these hazards. While many of the data bases and analysis methods exist for an adequate estimate of threat and plant response, this knowledge is not fully used and no comprehensive guidance has been developed. Siting of nuclear power plants relative to offsite hazardous materials is a risk based procedure that considers both probabilities and consequences of events that make up accident scenarios. In this context it appears feasible to improve the procedures vis-a-vis the perception of safety, economy of effort, and efficiency of implementation. A scenario dependent conditional risk approach is outlined as a possible means of improving the siting procedures.

  14. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    SciTech Connect (OSTI)

    Vigil, R.A. [Science & Engineering Associates, Inc., Albuquerque, NM (United States); Jacobus, M.J. [Sandia National Labs., Albuquerque, NM (United States)

    1994-04-01T23:59:59.000Z

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent.

  15. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect (OSTI)

    Rebak, Raul B. [General Electric] (ORCID:0000000280704475)

    2014-12-30T23:59:59.000Z

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding material both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

  16. Multi-view kernel construction

    E-Print Network [OSTI]

    Sa, Virginia R.; Gallagher, Patrick W.; Lewis, Joshua M.; Malave, Vicente L.

    2010-01-01T23:59:59.000Z

    5157-z Multi-view kernel construction Virginia R. de Sa ·multiple different graph construction algorithms. The Ng et

  17. Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

    2014-09-01T23:59:59.000Z

    The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

  18. Cladding embrittlement during postulated loss-of-coolant accidents.

    SciTech Connect (OSTI)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31T23:59:59.000Z

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  19. The Accident at Fukushima: What Happened?

    SciTech Connect (OSTI)

    Fujie, Takao [Japan Nuclear Technology Institute - JANTI (Japan)

    2012-07-01T23:59:59.000Z

    At 2:46 PM, on the coast of the Pacific Ocean in eastern Japan, people were spending an ordinary afternoon. The earthquake had a magnitude of 9.0, the fourth largest ever recorded in the world. Avery large number of aftershocks were felt after the initial earthquake. More than 100 of them had a magnitude of over 6.0. There were very few injured or dead at this point. The large earthquake caused by this enormous crustal deformation spawned a rare and enormous tsunami that crashed down 30-40 minutes later. It easily cleared the high levees, washing away cars and houses and swallowing buildings of up to three stories in height. The largest tsunami reading taken from all regions was 40 meters in height. This tsunami reached the West Coast of the United States and the Pacific coast of South America, with wave heights of over two meters. It was due to this tsunami that the disaster became one of a not imaginable scale, which saw the number of dead or missing reach about 20,000 persons. The enormous tsunami headed for 15 nuclear power plants on the Pacific coast, but 11 power plants withstood the tsunami and attained cold shutdown. The flood height of the tsunami that struck each power station ranged to a maximum of 15 meters. The Fukushima Daiichi Nuclear Power Plant Units experienced the largest and the cores of three reactors suffered meltdown. As a result, more than 160,000 residents were forced to evacuate, and are still living in temporary accommodation. The main focus of this presentation is on what happened at the Fukushima Daiichi, and how station personnel responded to the accident, with considerable international support. A year after the Fukushima Daiichi accident, Japan is in the process of leveraging the lessons learned from the accident to further improve the safety of nuclear power facilities and regain the trust of society. In this connection, not only international organizations, including IAEA, and WANO, but also governmental organizations and nuclear industry representatives from various countries, have been evaluating what happened at Fukushima Daiichi. Support from many countries has contributed to successfully stabilizing the Fukushima Daiichi Nuclear Power Station. International cooperation is required as Japan started along the long road to decommissioning the reactors. Such cooperation with the international community would achieve the decommissioning of the damaged reactors. Finally, recovery plans by the Japanese government to decontaminate surrounding regions have been started in order to get residents back to their homes as early as possible. Looking at the world's nuclear power industry, there are currently approximately 440 reactors in operation and 60 under construction. Despite the dramatic consequences of the Fukushima Daiichi catastrophe it is expected that the importance of nuclear power generation will not change in the years to come. Newly accumulated knowledge and capabilities must be passed on to the next generation. This is the duty put upon us and which is one that we must embrace.

  20. TotalView Training

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del SolStrengthening a solidSynthesisAppliances » Top InnovativeTopoisomeraseTotalView

  1. Risk communication with Fukushima residents affected by the Fukushima Daiichi accident at whole-body counting

    SciTech Connect (OSTI)

    Gunji, I.; Furuno, A.; Yonezawa, R.; Sugiyama, K. [Risk Communication Study Office, Japan Atomic Energy Agency 4-33 Muramatsu, Tokai-mura, Ibaraki, 319-1194 (Japan)

    2013-07-01T23:59:59.000Z

    After the Tokyo Electric Power Company (TEPCO) Fukushima Daiichi nuclear power plant accident, the Tokai Research and Development Center of the Japan Atomic Energy Agency (JAEA) have had direct dialogue as risk communication with Fukushima residents who underwent whole-body counting examination (WBC). The purpose of the risk communication was to exchange information and opinions about radiation in order to mitigate Fukushima residents' anxiety and stress. Two kinds of opinion surveys were performed: one survey evaluated residents' views of the nuclear accident itself and the second survey evaluated the management of WBC examination as well as the quality of JAEA's communication skills on risks. It appears that most Fukushima residents seem to have reduced their anxiety level after the direct dialogue. The results of the surveys show that Fukushima residents have the deepest anxiety and concern about their long-term health issues and that they harbor anger toward the government and TEPCO. On the other hand, many WBC patients and patients' relatives have expressed gratitude for help in reducing their feelings of anxiety.

  2. CFD modeling of debris melting phenomena during late phase Candu 6 severe accident

    SciTech Connect (OSTI)

    Nicolici, S.; Dupleac, D.; Prisecaru, I. [Univ. Politehnica of Bucharest, 313 Splaiul Independentei, 060042, Bucharest (Romania)

    2012-07-01T23:59:59.000Z

    The objective of this paper was to study the phase change of the debris formed on the Candu 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys-Fluent code. The 3D model using Large Eddy Simulation (LES) predicts the conjugate, radiative and convective heat transfer inside and from the corium pool. LES simulations require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. The preliminary results include heat transfer coefficients, temperature profiles and heat fluxes through calandria wall. From the safety point of view it is very important to maintain a heat flux through the wall below the CHF assuring the integrity of the calandria vessel. This can be achieved by proper cooling of the tank water which contains the vessel. Also, transient duration can be estimated being important in developing guidelines for severe accidents management. The debris physical structure and material properties have large uncertainties in the temperature range of interest. Thus, further sensitivity studies should be carried out in order to better understand the influence of these parameters on this complex phenomenon. (authors)

  3. Protective laser beam viewing device

    DOE Patents [OSTI]

    Neil, George R.; Jordan, Kevin Carl

    2012-12-18T23:59:59.000Z

    A protective laser beam viewing system or device including a camera selectively sensitive to laser light wavelengths and a viewing screen receiving images from the laser sensitive camera. According to a preferred embodiment of the invention, the camera is worn on the head of the user or incorporated into a goggle-type viewing display so that it is always aimed at the area of viewing interest to the user and the viewing screen is incorporated into a video display worn as goggles over the eyes of the user.

  4. Microsoft Word - 2015.06.22 - Report to Congress - Accident Tolerant...

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

    ROADMAP: DEVELOPMENT OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Development of Light Water...

  5. Analysing Aviation Accidents using WB-Analysis An Application for Multimodal Reasoning

    E-Print Network [OSTI]

    Moeller, Ralf

    Analysing Aviation Accidents using WB-Analysis An Application://www.rvs.uni-bielefeld.de We describe our ongoing work in accident analysis. Accident reports should tell* * us at least what the accident was and what the critical events were. A third requirement th* *ey should fulfil is to explain

  6. RESEARCH FOUNDATION -STATE UNIVERSITY OF NEW YORK REPORT OF ACCIDENT OR INJURY

    E-Print Network [OSTI]

    Suzuki, Masatsugu

    RESEARCH FOUNDATION - STATE UNIVERSITY OF NEW YORK REPORT OF ACCIDENT OR INJURY (OTHER THAN A MOTOR VEHICLE ACCIDENT) Revised: July 2008 1. Date and T ime of accident: Date: T ime: 2. Date of Report: 3. T o be completed by Safety Supervisor: YEAR: NO.: SEQUENCE: FILE ID: 4. Did accident involve personal injury? Yes

  7. INTRODUCTION OF FREQUENCY IN FRANCE FOLLOWING THE AZF ACCIDENT Clment LENOBLE*

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    INTRODUCTION OF FREQUENCY IN FRANCE FOLLOWING THE AZF ACCIDENT Clément LENOBLE* , Clarisse DURAND** * INERIS, Accident risks division, Parc Technologique Alata BP2, F-60550 Verneuil-en-Halatte ** French been consecutive to industrial accidents. Two years after the industrial accident of AZF (French

  8. Monthly Theme OARS January 2009 Report an Accident / Incident / Near Miss

    E-Print Network [OSTI]

    Calgary, University of

    Monthly Theme ­ OARS ­ January 2009 Report an Accident / Incident / Near Miss Online Accident Reporting System (OARS) debuts January 2009 EH&S has a NEW online system to report any accident or incident that happens at the University. The web- based reporting system is called OARS -- Online Accident Reporting

  9. Accidents, engineering and history at NASA: 1967-2003

    E-Print Network [OSTI]

    Brown, Alexander F. G. (Alexander Frederic Garder), 1970-

    2009-01-01T23:59:59.000Z

    The manned spaceflight program of the National Aeronautics and Space Administration (NASA) has suffered three fatal accidents: one in the Apollo program and two in the Space Transportation System (the Shuttle). These were ...

  10. accident hydrologic analysis: Topics by E-print Network

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

    (SFHS) is a non information, contact: - Neil JohnsonMWH - Jayantha ObeysekeraSFWMD - Mike SukopFIU - Chris PetersCH2M HILL Sukop, Mike 291 HOW TO REPORT AN ACCIDENT,...

  11. Type B Accident Investigation Board Report on the Head Injury...

    Office of Environmental Management (EM)

    on the Head Injury to a Miner at the Waste Isolation Pilot Plant, Carlsbad, New Mexico - August 25, 2004 Type B Accident Investigation Board Report on the Head Injury to a Miner at...

  12. Type B Accident Investigation Of The February 25, 2009 Injury...

    Energy Savers [EERE]

    To A Passenger In An Electric Cart At The Waste Isolation Pilot Plant, Carlsbad, New Mexico Type B Accident Investigation Of The February 25, 2009 Injury To A Passenger In An...

  13. accident consequence code: Topics by E-print Network

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

    MACPISA-CANDU (more) Pohl, Daniel J. 2009-01-01 5 Validation of severe accident codes against Phebus FP for plant applications: status of the PHEBEN2 project CiteSeer...

  14. A STAMP model of the Überlingen aircraft collision accident

    E-Print Network [OSTI]

    Wong, Brian, 1982 Nov 11-

    2004-01-01T23:59:59.000Z

    STAMP is a method for evaluating accidents that is based on systems theory. It departs from traditional event chain models that tend to focus on human errors instead of the goals and motives that triggered the errors. The ...

  15. Modeling control room crews for accident sequence analysis

    E-Print Network [OSTI]

    Huang, Y. (Yuhao)

    1991-01-01T23:59:59.000Z

    This report describes a systems-based operating crew model designed to simulate the behavior of an nuclear power plant control room crew during an accident scenario. This model can lead to an improved treatment of potential ...

  16. accidents traffic: Topics by E-print Network

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

    specifically, three distinct studies (more) OShields, Lara Lynn 2007-01-01 2 Does Daylight Savings Time Affect Traffic Accidents? Texas A&M University - TxSpace Summary: This...

  17. Type B Accident Investigation of the January 10, 2006, Flash...

    Office of Environmental Management (EM)

    January 10, 2006, Flash Fire and Injury at the Savannah River National Laboratory Type B Accident Investigation of the January 10, 2006, Flash Fire and Injury at the Savannah River...

  18. Response of Soviet VVER-440 accident localization systems to overpressurization

    SciTech Connect (OSTI)

    Kulak, R.F.; Fiala, C.; Sienicki, J.J.

    1989-01-01T23:59:59.000Z

    The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. Descriptions of the VVER accident localization structures may be found in the report DOE NE-0084. The objective of this paper is to evaluate the structural integrity of the VVER-440 ALS at the Soviet design pressure, and to determine their response to pressure loadings beyond the design value. Complex, three-dimensional, nonlinear, finite element models were developed to represent the major structural components of the localization systems of the VVER-440 models V230 and V213. The interior boundary of the localization system was incrementally pressurized in the calculations until the prediction of gross failure. 6 refs., 9 figs.

  19. Accident Investigation of the February 5, 2014, Underground Salt...

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

    5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the...

  20. Type B Accident Investigation Board Report of the Savannah River...

    Office of Environmental Management (EM)

    Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009 Type B Accident Investigation Board Report of the Savannah River Site Hand Injury at the...

  1. Type B Accident Investigation of the Savannah River Site Arc...

    Energy Savers [EERE]

    H2 Demolition, in Niskayuna, New, York Type B Accident Investigation Board Report of the Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009...

  2. Failsafe : living with man-made disaster and accident

    E-Print Network [OSTI]

    Higgins, Saoirse, 1966-

    2004-01-01T23:59:59.000Z

    "There is no progress with out progress of the catastrophe." Virilio. This thesis project proposes that technological solutions in the design of our systems are not enough to prevent 'man-made' accident. Social, organisational ...

  3. accidents radiologiques sur: Topics by E-print Network

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

    SUR L'ORIGINE DES BRUITS DU COEUR ET DES ACCIDENTS DE DCOMPRESSION EN PLONGE : LA CAVITATION. Engineering Websites Summary: CAVITATION. HYPOTH?SE SUR L'ORIGINE DE LA FERMETURE...

  4. The 2011 Tohoku earthquake, tsunami, and Fukushima nuclear accident

    E-Print Network [OSTI]

    Ferrari, Silvia

    The 2011 Tohoku earthquake, tsunami, and Fukushima nuclear accident: the Risk Policy Aftermath 3 #12;Personal experience in March 2011 Tsukuba 170km Tokyo 230km Fukushima Daiichi nuclear power

  5. Type B Accident Investigation of the Acid Vapor Inhalation on...

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

    2005, in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory Type B Accident Investigation of the Acid Vapor Inhalation on June 7, 2005, in TA-48, Building RC-1...

  6. accident locations: Topics by E-print Network

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

    17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Exact Location : Date of Accident : AM PM Environmental Sciences and Ecology Websites Summary: SSN Cell Phone Home...

  7. accidents dus aux: Topics by E-print Network

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

    qui taient en lgre hausse lanne dernire. De plus, il ny a eu aucun accident de trajet avec arrt en 2002. Pour le personnel LHC gnie civil, les rsultats...

  8. accident risks methodology: Topics by E-print Network

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

    13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design Basis Accident Radiological Assessment Calculational Methodology CiteSeer Summary: submitted revised...

  9. Type B Accident Investigation Board Report on the September 1...

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

    September 1, 1999, Plutonium Intakes at the Savannah River Site FB-Line Type B Accident Investigation Board Report on the September 1, 1999, Plutonium Intakes at the Savannah River...

  10. accident phebus program: Topics by E-print Network

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

    Test FPT1 CiteSeer Summary: The contribution of radioiodine to risk from a severe accident is recognized to be one of the highest among all the fission products. In a long term...

  11. Some methods of estimating uncertainty in accident reconstruction

    E-Print Network [OSTI]

    Milan Batista

    2011-07-20T23:59:59.000Z

    In the paper four methods for estimating uncertainty in accident reconstruction are discussed: total differential method, extreme values method, Gauss statistical method, and Monte Carlo simulation method. The methods are described and the program solutions are given.

  12. accident dosimetry system: Topics by E-print Network

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

    11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 A New Accident Model for Engineering Safer Systems Nancy Leveson Engineering Websites Summary: A New...

  13. Type B Accident Investigation Board Report Employee Puncture...

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

    F-TRU Waste Remediation Facility at the Savannah River Site on June 14, 2010 Type B Accident Investigation Board Report Employee Puncture Wound at the F-TRU Waste Remediation...

  14. Iodine chemical forms in LWR severe accidents

    SciTech Connect (OSTI)

    Beahm, E.C.; Weber, C.F.; Kress, T.S.; Parker, G.W.

    1991-01-01T23:59:59.000Z

    Calculated data from seven severe accident sequences in light-water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I{sup {minus}}). The radiation induced conversion of I{sup {minus}} in water pools into I{sub 2} is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled, and large fractions of iodine as I{sub 2} within the containment atmosphere may be produced. 16 refs.

  15. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01T23:59:59.000Z

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  16. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01T23:59:59.000Z

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  17. Evaluation of accident frequencies at the canister storage building

    SciTech Connect (OSTI)

    LIU, Y.J.

    1999-05-13T23:59:59.000Z

    By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multi-canister overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

  18. Trees as Filters of Radioactive Fallout from the Chernobyl Accident

    E-Print Network [OSTI]

    Brownridge, James D

    2011-01-01T23:59:59.000Z

    This paper is a copy of an unpublished study of the filtering effect of red maple trees (acer rubrum) on fission product fallout near Binghamton, NY, USA following the 1986 Chernobyl accident. The conclusions of this work may offer some insight into what is happening in the forests exposed to fallout from the Fukushima Daiichi Nuclear Plant accident. This posting is in memory of Noel K. Yeh.

  19. Berkeley Lab Accident Statistics Through July 31, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through July 31, 2009 These slides are updated on a monthly Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1.92 3.90 3.41 2.65 2

  20. Berkeley Lab Accident Statistics Through September 30, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through September 30, 2009 These slides are updated on a monthly Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1.92 3.90 3.41 2.65 2

  1. Berkeley Lab Accident Statistics Through May 31, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through May 31, 2009 These slides are updated on a monthly basis Berkeley Lab Site Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1

  2. Berkeley Lab Accident Statistics Through October 31, 2008

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through October 31, 2008 These slides are updated on a monthly Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1.72 0.40 3.90 3.41 2

  3. Berkeley Lab Accident Statistics Through August 31, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    1 Berkeley Lab Accident Statistics Through August 31, 2009 These slides are updated on a monthly Accident Rates 5.70 4.95 3.79 2.92 2.93 3.27 3.63 2.44 2.17 2.51 1.17 1.81 1.28 1.65 1.92 3.90 3.41 2.65 2

  4. APT Blanket System Loss-of-Helium-Gas Accident Based on Initial Conceptual Design - Helium Supply Rupture into Blanket Module

    SciTech Connect (OSTI)

    Hamm, L.L.

    1998-10-07T23:59:59.000Z

    The model results are used to determine if beam power shutdown is necessary (or not) as a result of the LOHGA accident to maintain the blanket system well below any of the thermal-hydraulic constraints imposed on the design. The results also provide boundary conditions to the detailed bin model to study the detailed temperature response of the hot blanket module structure. The results for these two cases are documented in the report.

  5. Type B Accident Investigation Board Report on the March 27, 1998, Rotating Shaft Accident at the Ames Laboratory, Ames, Iowa

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by John Kennedy, Acting Manager, Chicago Operations Office, U.S. Department of Energy (DOE).

  6. False color viewing device

    DOE Patents [OSTI]

    Kronberg, J.W.

    1991-05-08T23:59:59.000Z

    This invention consists of a viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching, the user`s eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage.

  7. False color viewing device

    DOE Patents [OSTI]

    Kronberg, James W. (108 Independent Blvd., Aiken, SC 29801)

    1992-01-01T23:59:59.000Z

    A viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching the user's eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage.

  8. False color viewing device

    DOE Patents [OSTI]

    Kronberg, J.W.

    1992-10-20T23:59:59.000Z

    A viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching the user's eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage. 7 figs.

  9. The view from Kiev

    SciTech Connect (OSTI)

    Kiselyov, S.

    1993-11-01T23:59:59.000Z

    This article reports the observations of correspondents for the Bulletin (two Russian journalists, one based in Moscow, the other in Kiev) who investigated the status of the Soviet Union's Black Sea Fleet and Ukraine's status as a non-nuclear-weapons state. After two years of wrangling and two earlier failed settlements, Russian President Boris Yeltsin met with Ukrainian President Leonid Kravchuk at Massandra in Crimea. On September 3, the leaders announced that Russia would buy out Ukraine's interest in the fleet and lease the port at Sevastopol. The Massandra summit was also supposed to settle Ukraine's status as a non-nuclear-weapons state. Described here are the Kiev-based correspondent's views on the Massandra summit (and its major topics), which was to have been called off by the Russian foreign ministry when Ukrainian Prime Minister Leonid Kuchma resigned.

  10. Civil aircraft accident Report on the Accident to Boeing 707-465 G-Arwe at Heathrow Airport, London on 8th April 1968 

    E-Print Network [OSTI]

    Anonymous

    1969-01-01T23:59:59.000Z

    A3.C.A.P. 324 Civil aircraft accident Report on the Accident to Boeing 707-465 G-Arwe at Heathrow Airport, London on 8th April 1968...

  11. Alcohol, Drugs, and Accident Prevention (RC-371/-571) Course Description The role of alcohol and drugs and their relationship to accident causation will be examined. The problem

    E-Print Network [OSTI]

    Wu, Mingshen

    Alcohol, Drugs, and Accident Prevention (RC-371/-571) Course Description The role of alcohol and drugs and their relationship to accident causation will be examined. The problem of alcoholism and drug

  12. NORMES D'ACTUACI EN CAS D'ACCIDENT 1. Els accidents hauran de justificar-se mitjanant la corresponent comunicaci

    E-Print Network [OSTI]

    Geffner, Hector

    NORMES D'ACTUACI� EN CAS D'ACCIDENT 1. Els accidents hauran de justificar-se mitjançant la corresponent comunicació d'accident que haurà d'emplenar el club, entitat esportiva o empresa prenedora a la als serveis mèdics concertats és necessari aportar la comunicació d'accident certificada i identificar

  13. Initial Cladding Condition

    SciTech Connect (OSTI)

    E. Siegmann

    2000-08-22T23:59:59.000Z

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in evaluating the post-closure performance of the Monitored Geologic Repository (MGR) in relation to waste form degradation.

  14. View dependent fluid dynamics

    E-Print Network [OSTI]

    Barran, Brian Arthur

    2006-08-16T23:59:59.000Z

    , are modified to support a nonuniform simulation grid. In addition, infinite fluid boundary conditions are introduced that allow fluid to flow freely into or out of the simulation domain to achieve the effect of large, boundary free bodies of fluid. Finally, a...

  15. Human factors review for Severe Accident Sequence Analysis (SASA)

    SciTech Connect (OSTI)

    Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

    1984-01-01T23:59:59.000Z

    The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure.

  16. LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN E. A. Mogahed, L. El-Guebaly, A. Abdou, P. Wilson, D. Henderson and the ARIES Team

    E-Print Network [OSTI]

    California at San Diego, University of

    LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN E. A. Mogahed, L accident (LOCA) and loss of flow accident (LOFA) analysis is performed for ARIES-AT, an advanced fusion of steel in the reactor is about (600 °C - 700°C) after about 4 days from the onset of the accident

  17. Security Conditions

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

    2004-07-08T23:59:59.000Z

    This Notice ensures that DOE uniformly meets the requirements of the Homeland Security Advisory System outlined in Homeland Security Presidential Directive-3, Threat Conditions and Associated Protective Measures, dated 3-11-02, and provides responses specified in Presidential Decision Directive 39, U.S. Policy on Counterterrorism (U), dated 6-21-95. It cancels DOE N 473.8, Security Conditions, dated 8-7-02. Extended until 7-7-06 by DOE N 251.64, dated 7-7-05 Cancels DOE N 473.8

  18. Managing Errors to Reduce Accidents in High Consequence Networked Information Systems

    SciTech Connect (OSTI)

    Ganter, J.H.

    1999-02-01T23:59:59.000Z

    Computers have always helped to amplify and propagate errors made by people. The emergence of Networked Information Systems (NISs), which allow people and systems to quickly interact worldwide, has made understanding and minimizing human error more critical. This paper applies concepts from system safety to analyze how hazards (from hackers to power disruptions) penetrate NIS defenses (e.g., firewalls and operating systems) to cause accidents. Such events usually result from both active, easily identified failures and more subtle latent conditions that have resided in the system for long periods. Both active failures and latent conditions result from human errors. We classify these into several types (slips, lapses, mistakes, etc.) and provide NIS examples of how they occur. Next we examine error minimization throughout the NIS lifecycle, from design through operation to reengineering. At each stage, steps can be taken to minimize the occurrence and effects of human errors. These include defensive design philosophies, architectural patterns to guide developers, and collaborative design that incorporates operational experiences and surprises into design efforts. We conclude by looking at three aspects of NISs that will cause continuing challenges in error and accident management: immaturity of the industry, limited risk perception, and resource tradeoffs.

  19. Improvement design study on steam generator of MHR-50/100 aiming higher safety level after water ingress accident

    SciTech Connect (OSTI)

    Oyama, S. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan); Minatsuki, I.; Shimizu, K. [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan)

    2012-07-01T23:59:59.000Z

    Mitsubishi Heavy Industries, Ltd. (MHI) has been studying on MHI original High Temperature Gas cooled Reactor (HTGR), namely MHR-50/100, for commercialization with supported by JAEA. In the heat transfer system, steam generator (SG) is one of the most important components because it should be imposed a function of heat transfer from reactor power to steam turbine system and maintaining a nuclear grade boundary. Then we especially focused an effort of a design study on the SG having robustness against water ingress accident based on our design experience of PWR, FBR and HTGR. In this study, we carried out a sensitivity analysis from the view point of economic and plant efficiency. As a result, the SG design parameter of helium inlet/outlet temperature of 750 deg. C/300 deg. C, a side-by-side layout and one unit of SG attached to a reactor were selected. In the next, a design improvement of SG was carried out from the view point of securing the level of inherent safety without reliance on active steam dump system during water ingress accident considering the situation of the Fukushima nuclear power plant disaster on March 11, 2011. Finally, according to above basic design requirement to SG, we performed a conceptual design on adapting themes of SG structure improvement. (authors)

  20. Methods for air cleaning system design and accident analysis

    SciTech Connect (OSTI)

    Gregory, W.S.; Nichols, B.D.

    1986-01-01T23:59:59.000Z

    This paper describes methods, in the form of a handbook and five computer codes, that can be used for air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes, they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes' availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described.

  1. PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise

    SciTech Connect (OSTI)

    Hill, Robin L.; Conrady, Matthew M.

    2011-10-28T23:59:59.000Z

    This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

  2. Accident Sequence Evaluation Program: Human reliability analysis procedure

    SciTech Connect (OSTI)

    Swain, A.D.

    1987-02-01T23:59:59.000Z

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  3. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect (OSTI)

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K. [Reactor Safety and Analysis, Nuclear Power Corporation of India Ltd., Nabhkiya Urja Bhavan, Anushakti Nagar, Mumbai Maharashtra 400094 (India)

    2012-07-01T23:59:59.000Z

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  4. FSAR fire accident analysis for a plutonium facility

    SciTech Connect (OSTI)

    Lam, K.

    1997-06-01T23:59:59.000Z

    The Final Safety Analysis Report (FSAR) for a plutonium facility as required by DOE Orders 5480.23 and 5480.22 has recently been completed and approved. The facility processes and stores radionuclides such as Pu-238, Pu-239, enriched uranium, and to a lesser degree other actinides. This facility produces heat sources. DOE Order 5480.23 and DOE-STD-3009-94 require analysis of different types of accidents (operational accidents such as fires, explosions, spills, criticality events, and natural phenomena such as earthquakes). The accidents that were analyzed quantitatively, or the Evaluation Basis Accidents (EBAs), were selected based on a multi-step screening process that utilizes extensively the Hazards Analysis (HA) performed for the facility. In the HA, specific accident scenarios, with estimated frequency and consequences, were developed for each identified hazard associated with facility operations and activities. Analysis of the EBAs and comparison of their consequences to the evaluation guidelines established the safety envelope for the facility and identified the safety-class structures, systems, and components. This paper discusses the analysis of the fire EBA. This fire accident was analyzed in relatively great detail in the FSAR because of its potential off-site consequences are more severe compared to other events. In the following, a description of the scenario is first given, followed by a brief summary of the methodology for calculating the source term. Finally, the author discuss how a key parameter affecting the source term, the leakpath factor, was determined, which is the focus of this paper.

  5. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    SciTech Connect (OSTI)

    Siefken, Larry James

    1999-02-01T23:59:59.000Z

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

  6. Effect of shape reactivity on the rod-ejection accident

    SciTech Connect (OSTI)

    Neogy, P.; Carew, J.F.

    1982-09-01T23:59:59.000Z

    The shape reactivity has a significant influence on the rod ejection accident. After the control rod is fully ejected from the core, the neutron flux undergoes a large reduction at the ejected rod location. The corresponding effect on the control reactivity is comparable in magnitude to the Doppler reactivity, and makes a significant contribution to limiting the power excursion during the transient. The neglect of this effect in point kinetics and space time synthesis analyses of the rod ejection accident may account in part for the large degree of conservatism usually associated with these analyses.

  7. Radionuclide-inventory impacts on reactor-accident consequences

    SciTech Connect (OSTI)

    Ostmeyer, R.M.

    1981-01-01T23:59:59.000Z

    To examine the potential impacts of the different radionuclide inventories on predicted accident consequences, and the appropriateness of inventory scaling, a series of calculations was performed using CRAC2, a modified version of the WASH-1400 consequence model. Consequences were calculated assuming (1) an SST-1 accident (large scale core melt with uncontrolled release directly to the atmosphere), (2) Indian Point population and wind-rose data, (3) New York City weather data, and (4) a distribution of evacuations within 16 km of the reactor.

  8. Improved assessment of population doses and risk factors for a nuclear power plant under accident conditions

    E-Print Network [OSTI]

    Meyer, Christopher Martin

    1985-01-01T23:59:59.000Z

    considered No. Radionuclide Core Inventory (Curies x 1. 0 E-08) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 Co-58 Co-60 Kr-85... classifications: precipitation scavenging or wet deposition and dry deposition. 21 Precipitation scavenging of airborne materials can occur by in- cloud or by below-cloud processes. In-cloud scavenging involves condensation precipitation in which the airborne...

  9. Utilizing an encroachment probability benefit-cost model to estimate accident reduction factors

    E-Print Network [OSTI]

    Hayes, Carolyn A

    1997-01-01T23:59:59.000Z

    Improving safety on Texas roadways is a major public concern. Over the years, the Texas Department of Transportation and other highway agencies have become interested in reducing society's accident cost while maximizing returns on accident...

  10. The Effect of the 18-Year Old Drinking Age on Auto Accidents

    E-Print Network [OSTI]

    Cucchiaro, Stephen

    The effect of Massachusetts' reduced drinking age on auto accidents is examined by employing an interrupted time series analysis of monthly accident data covering the period January, 1969, through September 1973. The data ...

  11. STAMP-Based Analysis of a Refinery Overflow Accident Nancy Leveson, Margaret Stringfellow, and John Thomas

    E-Print Network [OSTI]

    Leveson, Nancy

    1 STAMP-Based Analysis of a Refinery Overflow Accident Nancy Leveson, Margaret Stringfellow, and John Thomas As an example of STAMP, we have taken an accident report produced for a real refinery

  12. accident efectele medico-biologice: Topics by E-print Network

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

    Phone Home Phone Work Phone Exact Location : Date of Accident : AM PM Date accident treatment provided? Yes No Where Was time lost from work? Yes No If yes, how long? Could this...

  13. Three dimensional effects in analysis of PWR steam line break accident

    E-Print Network [OSTI]

    Tsai, Chon-Kwo

    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) ...

  14. Calculation Notes for Subsurface Leak Resulting in Pool, TWRS FSAR Accident Analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25T23:59:59.000Z

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Subsurface Leaks Resulting in Pool.

  15. Calculation notes for surface leak resulting in pool, TWRS FSAR accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25T23:59:59.000Z

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Surface Leaks Resulting in Pool.

  16. Natural Language Processing (NLP) tools for the analysis of incident and accident reports

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Natural Language Processing (NLP) tools for the analysis of incident and accident reports, Analysis of accidents/incidents, Categorization, Textual similarity INTRODUCTION Learning valuable lessons, (ii) the analysis of reports regardless of the categorization in order to expand the analysis

  17. Accident Analysis and Prevention 42 (2010) 213224 Contents lists available at ScienceDirect

    E-Print Network [OSTI]

    Wisconsin at Madison, University of

    Accident Analysis and Prevention 42 (2010) 213­224 Contents lists available at ScienceDirect Accident Analysis and Prevention journal homepage: www.elsevier.com/locate/aap Rainfall effect on single

  18. Type A Accident Investigation Board Report on the July 1, 2008...

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

    July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration Type A Accident Investigation Board Report on the July 1, 2008, of the Vehicle Fatality...

  19. University of Virginia Agency 207 Accident Report for Workers' Compensation Claim Please complete this form and turn it in to your department's Human Resource Coordinator or

    E-Print Network [OSTI]

    Acton, Scott

    University of Virginia Agency 207 Accident Report for Workers' Compensation Claim Please complete Accident Reported: __________Reported Accident to:___________________________________ Was Supervisor(es)___________________________________________________________________ Information About the Nature and Cause of Accident Machine, tool, or object causing injury

  20. A Regulator's View of Cogeneration

    E-Print Network [OSTI]

    Shanaman, S. M.

    1982-01-01T23:59:59.000Z

    of the total national electric generation. In view of the energy requirements of Pennsylvania's industry and the impact of increasing energy costs on employment the Commission directed its technical staff to investigate the potential for industrial cogeneration...

  1. Development of integrated core disruptive accident analysis code for FBR - ASTERIA-FBR

    SciTech Connect (OSTI)

    Ishizu, T.; Endo, H.; Tatewaki, I.; Yamamoto, T. [Japan Nuclear Energy Safety Organization JNES, Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo (Japan); Shirakawa, N. [Inst. of Applied Energy IAE, Shimbashi SY Bldg., 14-2 Nishi-Shimbashi 1-Chome, Minato-ku, Tokyo (Japan)

    2012-07-01T23:59:59.000Z

    The evaluation of consequence at the severe accident is the most important as a safety licensing issue for the reactor core of liquid metal cooled fast breeder reactor (LMFBR), since the LMFBR core is not in an optimum condition from the viewpoint of reactivity. This characteristics might induce a super-prompt criticality due to the core geometry change during the core disruptive accident (CDA). The previous CDA analysis codes have been modeled in plural phases dependent on the mechanism driving a super-prompt criticality. Then, the following event is calculated by connecting different codes. This scheme, however, should introduce uncertainty and/or arbitrary to calculation results. To resolve the issues and obtain the consistent calculation results without arbitrary, JNES is developing the ASTERIA-FBR code for the purpose of providing the cross-check analysis code, which is another required scheme to confirm the validity of the evaluation results prepared by applicants, in the safety licensing procedure of the planned high performance core of Monju. ASTERIA-FBR consists of the three major calculation modules, CONCORD, dynamic-GMVP, and FEMAXI-FBR. CONCORD is a three-dimensional thermal-hydraulics calculation module with multi-phase, multi-component, and multi-velocity field model. Dynamic-GMVP is a space-time neutronics calculation module. FEMAXI-FBR calculates the fuel pellet deformation behavior and fuel pin failure behavior. This paper describes the needs of ASTERIA-FBR development, major module outlines, and the model validation status. (authors)

  2. Radiological Release Accident Investigation Report (Phase II Report)

    Broader source: Energy.gov [DOE]

    In its Phase II Report, the Accident Investigation Board concludes that the Feb. 14, 2014 radiological release at the Waste Isolation Pilot Plant was caused by an exothermic reaction involving the mixture of organic materials and nitrate salts in one drum that was processed at the Los Alamos National Laboratory in December 2013.

  3. accident codes applications: Topics by E-print Network

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

    accident codes applications First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Validation of severe...

  4. Accident Research Helps Save Lives of Loggers Research Brief # 33

    E-Print Network [OSTI]

    of the Strategic Partnership Agreement between logging companies and the federal Occupational Safety & Health Administration (OSHA), 257 logging companies agreed to submit their accident reports annually to the Louisiana and continues to present the results of new logging safety research to nearly every logging company

  5. alamos nuclear accidents: Topics by E-print Network

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

    alamos nuclear accidents First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Fukushima Nuclear Power Plant...

  6. accident nuclear materials: Topics by E-print Network

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

    accident nuclear materials First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NUREGCR-7034 Analysis of...

  7. Application of Evidential Networks in quantitative analysis of railway accidents

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Application of Evidential Networks in quantitative analysis of railway accidents Felipe Aguirre1 during the qualitative or quantitative evaluation of risk. Several new techniques for HRA were invented. As a consequence, safety engineers try to take into account this factor in risk assessment. However, human

  8. Berkeley Lab Accident Statistics Through December 31, 2009

    E-Print Network [OSTI]

    Eisen, Michael

    Engineering Environmental Energy Tech. Genomics Life Sciences Materials Sciences NERSC Center Nuclear Science Energy Tech. Genomics Life Sciences Materials Sciences NERSC Center Nuclear Science Physical Biosciences1 Berkeley Lab Accident Statistics Through December 31, 2009 These slides are updated on a monthly

  9. Source terms for plutonium aerosolization from nuclear weapon accidents

    SciTech Connect (OSTI)

    Stephens, D.R.

    1995-07-01T23:59:59.000Z

    The source term literature was reviewed to estimate aerosolized and respirable release fractions for accidents involving plutonium in high-explosive (HE) detonation and in fuel fires. For HE detonation, all estimates are based on the total amount of Pu. For fuel fires, all estimates are based on the amount of Pu oxidized. I based my estimates for HE detonation primarily upon the results from the Roller Coaster experiment. For hydrocarbon fuel fire oxidation of plutonium, I based lower bound values on laboratory experiments which represent accident scenarios with very little turbulence and updraft of a fire. Expected values for aerosolization were obtained from the Vixen A field tests, which represent a realistic case for modest turbulence and updraft, and for respirable fractions from some laboratory experiments involving large samples of Pu. Upper bound estimates for credible accidents are based on experiments involving combustion of molten plutonium droplets. In May of 1991 the DOE Pilot Safety Study Program established a group of experts to estimate the fractions of plutonium which would be aerosolized and respirable for certain nuclear weapon accident scenarios.

  10. Interdisciplinary Institute for Innovation Le risque d'accident nuclaire

    E-Print Network [OSTI]

    Boyer, Edmond

    . Cochran. (2011), Fukushima nuclear disaster and its implication for US nuclear power reactors. Ce chiffre Lévêque L'accident de Fukushima Daiichi s'est produit le 11 mars 2011. Cette catastrophe nucléaire irrémédiablement associée à une centrale nucléaire dont l'homme a perdu le contrôle. Fukushima Daiichi a ainsi fait

  11. Source: http://www.ftc.gov/debtcollection ost accidents

    E-Print Network [OSTI]

    Oliver, Douglas L.

    Source: http://www.ftc.gov/debtcollection M ost accidents on the job are caused by un- safe actions-852-4392 #12;Source: Office of Dietary Supplements, National Institutes of Health F requently the missing piece to finding work-life balance is "mindset." This is a psycho- logical leap to rock-solid commit- ment

  12. Accident Analysis and Prevention 40 (2008) 12441248 Short communication

    E-Print Network [OSTI]

    McLeod, Ian

    2008-01-01T23:59:59.000Z

    Accident Analysis and Prevention 40 (2008) 1244­1248 Short communication Power computations in time the intervention analysis of the policy, it is important to ensure that the statistical tests have enough power in a wide variety of traffic safety intervention analysis applications. Our method is illustrated

  13. accident sequence evaluation: Topics by E-print Network

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

    accident sequence evaluation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Resugaring: Lifting...

  14. accident research network: Topics by E-print Network

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

    accident research network First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Structure Evolution of...

  15. accident reports: Topics by E-print Network

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

    accident reports First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 TIPS ON ACCIDENTINCIDENT REPORTING...

  16. accident core heatup: Topics by E-print Network

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

    accident core heatup First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Appendix 10 Spent Fuel Heatup Time...

  17. accident management: Topics by E-print Network

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

    accident management First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Medical Management of Radiation...

  18. accident alarm systems: Topics by E-print Network

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

    accident alarm systems First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 FIRE ALARM SYSTEMS AND...

  19. accident research priorities: Topics by E-print Network

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

    accident research priorities First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Chapter 6: Research...

  20. accident rates: Topics by E-print Network

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

    accident rates First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Tangent length and sight distance...

  1. accident analysis software: Topics by E-print Network

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

    accident analysis software First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 The Role of Software in...

  2. accident and emergency medicine: Topics by E-print Network

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

    accident and emergency medicine First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Nanotechnology:...

  3. accident management measures: Topics by E-print Network

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

    accident management measures First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Medical Management of...

  4. accident retrospective estimation: Topics by E-print Network

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

    accident retrospective estimation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Some methods of...

  5. accident frequencies program: Topics by E-print Network

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

    accident frequencies program First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 INTRODUCTION OF FREQUENCY...

  6. advanced accident sequence: Topics by E-print Network

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

    advanced accident sequence First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Math 55a: Honors Advanced...

  7. accident consequence model: Topics by E-print Network

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

    accident consequence model First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Conservation consequences of...

  8. accident dinamika formirovaniya: Topics by E-print Network

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

    accident dinamika formirovaniya First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Adaptv dinamika: mirt l...

  9. accident recovery workers: Topics by E-print Network

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

    accident recovery workers First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 COLUMBIA UNIVERSITY...

  10. accident sequence analysis: Topics by E-print Network

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

    accident sequence analysis First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Biological Sequence Analysis...

  11. accident sequences simulated: Topics by E-print Network

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

    accident sequences simulated First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Combining Simulation with...

  12. accident reconstruction: Topics by E-print Network

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

    accident reconstruction First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Photogrammetry in Traffic...

  13. accident management epr: Topics by E-print Network

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

    accident management epr First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Medical Management of Radiation...

  14. accident alarm system: Topics by E-print Network

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

    accident alarm system First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 FIRE ALARM SYSTEMS AND PROCEDURES...

  15. accident progression event: Topics by E-print Network

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

    accident progression event First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 EVENT CALCULUS,...

  16. accident draft resolution: Topics by E-print Network

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

    accident draft resolution First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 CLASNote95Draft Spatial...

  17. accident dose consequences: Topics by E-print Network

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

    accident dose consequences First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Conservation consequences of...

  18. accident declaration form: Topics by E-print Network

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

    accident declaration form First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 DECLARATION FORM Page 1 of 2...

  19. accident release fractions: Topics by E-print Network

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

    accident release fractions First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 The GALEX Arecibo SDSS...

  20. accident excursion occurring: Topics by E-print Network

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

    accident excursion occurring First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Annual Report of Railroad...

  1. accident causes: Topics by E-print Network

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

    accident causes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Abstract--Petroleum transportation...

  2. accident aspectos vigentes: Topics by E-print Network

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

    accident aspectos vigentes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 6. Aspectos pendientes por...

  3. accident consequences otsenka: Topics by E-print Network

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

    accident consequences otsenka First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Conservation consequences...

  4. accident flying squad: Topics by E-print Network

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

    accident flying squad First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Governing the Mod Squad Walt...

  5. accident sequence quantification: Topics by E-print Network

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

    accident sequence quantification First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Quantification of the...

  6. accident consequence assessment: Topics by E-print Network

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

    accident consequence assessment First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Assessing the...

  7. accident consequence assessments: Topics by E-print Network

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

    accident consequence assessments First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Assessing the...

  8. accident medical aspect: Topics by E-print Network

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

    accident medical aspect First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 UNIVERSITY OF KENTUCKY REQUEST...

  9. accident tolerable continual: Topics by E-print Network

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

    accident tolerable continual First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Fault-Tolerant...

  10. accident records: Topics by E-print Network

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

    accident records First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 UMASS AMHERST POLICE DEPARTMENT PUBLIC...

  11. accident experimental facility: Topics by E-print Network

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

    accident experimental facility First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Project X Experimental...

  12. Berkeley Lab Accident Statistics Through August 31, 2010

    E-Print Network [OSTI]

    Eisen, Michael

    ) on walkway along McMillan Road. · Administrative Assistant ­ Puncture of finger ­ Received staple puncture1 Berkeley Lab Accident Statistics Through August 31, 2010 These slides are updated on a monthly;2 Narrative of August 2010 Recordable Injury Cases · Student Assistant ­ Strain/sprain of ankles, bruises

  13. Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident into the Pacific Ocean after the Fukushima accident and analysis of its dispersion in Japanese coastal waters, J into the ocean from the Fukushima Daiichi nuclear power plant (NPP) after the accident in March 2011 and to gain

  14. A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents

    E-Print Network [OSTI]

    Boyer, Edmond

    A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN , L Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN a , L. CANTREL a , C Accidents Majeurs (DPAM), CEN Cadarache - France 1 b Commissariat à l'Energie Atomique (CEA), Direction de l'Energie

  15. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    RITTMANN, P.D.

    1999-10-07T23:59:59.000Z

    Three bounding accidents postdated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing, and a hydrogen explosion. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  16. Development and use of the ESReDA directory of accident databases involving chemicals

    E-Print Network [OSTI]

    Boyer, Edmond

    , Data analysis, Reliability, Uncertainty, Accident, Hazardous material, Risk analysis. INTRODUCTION to the quality of data collected from accident investigations and their subsequent analysis. Many frameworks, a working group ofEuropean Safety, Reliability and Data Association (ESReDA) "Accident Analysis" (AA

  17. Accident Analysis and Prevention 42 (2010) 364371 Contents lists available at ScienceDirect

    E-Print Network [OSTI]

    Boggess, May M.

    2010-01-01T23:59:59.000Z

    Accident Analysis and Prevention 42 (2010) 364­371 Contents lists available at ScienceDirect Accident Analysis and Prevention journal homepage: www.elsevier.com/locate/aap Review Age-related safety Keywords: Age factors Truck Traffic accidents/crashes Occupation Safety Statistics a b s t r a c

  18. Presentation of progress of work in the "Accident Analysis" working group

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Presentation of progress of work in the "Accident Analysis" working group J.P. PINEAU INERIS Summary The "Accident Analysis" - AA - working group, initiated in January 1993, was at the origin of this investigation were presented at the Autumn 1994 ESReDA Seminar on Accident Analysis. A second step of the AA

  19. Biohazardous Laboratory Incidence/Accident Response and Reporting Protocol UGA Office of Biosafety

    E-Print Network [OSTI]

    Arnold, Jonathan

    Biohazardous Laboratory Incidence/Accident Response and Reporting Protocol UGA Office of Biosafety Biohazardous laboratory incident or accident involves the following: 1. Any potential or known exposure-related accidents or illnesses involving work described under the NIH Guidelines for Recombinant DNA Research (NIH

  20. Hazmat Accident Education -An Integrated Shirley E. Clark, Ph.D., P.E.

    E-Print Network [OSTI]

    Pitt, Robert E.

    Hazmat Accident Education - An Integrated Approach By Shirley E. Clark, Ph.D., P.E. Formerly Accident Education - An Integrated Approach 6. Performing Organization Code 7. Authors Shirley E. Clark, accidents, emergency planning, emergency response 18. Distribution Statement 19. Security Classification (of

  1. NO NAME:Accident reporting and Auto Insurance.doc July 15, 2014

    E-Print Network [OSTI]

    Kelly, Scott David

    NO NAME:Accident reporting and Auto Insurance.doc July 15, 2014 STATEMENT OF RESOURCES TO ADDRESS CLAIMS ARISING FROM ACCIDENTS INVOLVING VEHICLES OPERATED ON UNIVERSITY BUSINESS This statement contains a general description of resources available in connection with claims arising from accidents involving

  2. Accident Analysis and Prevention, 2012 (49), pp 73-77 www.elsevier.com/locate/aap

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Accident Analysis and Prevention, 2012 (49), pp 73-77 www.elsevier.com/locate/aap doi:10.1016/j.aap.2011.07.013 Motorcyclists' speed and "looked-but-failed-to-see" accidents Nicolas Clabaux, Thierry of accidents in which a non-priority road user failed to give way to an approaching motorcyclist without seeing

  3. Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment

    E-Print Network [OSTI]

    Del Moral , Pierre

    Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment #12;The SYSTEMS WITH APPLICATIONS TO ACCIDENT RISK ASSESSMENT DISSERTATION to obtain the doctor's degree promotor Prof. dr. A. Bagchi #12;Contents 1 Introduction 3 1.1 Accident risk assessment

  4. Extending the Borders of Accident Investigation: Applying Novel Analysis Techniques to the Loss of the Brazilian

    E-Print Network [OSTI]

    Johnson, Chris

    -1- Extending the Borders of Accident Investigation: Applying Novel Analysis Techniques to the Loss. In consequence, it is becoming increasingly difficult to identify the causes of incidents and accidents back to the development of a number of novel accident investigation techniques. Most of these approaches are intended

  5. DIRECTORY OF ESReDA ACCIDENT DATABASES Jean-Philippe PINEA

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    98-62 DIRECTORY OF ESReDA ACCIDENT DATABASES Jean-Philippe PINEAÃ? INERIS - Parc Technologique Alata, Reliability and Data Association (ESReDA) "Accident Analysis" has been dealing with data collection, quality, reliability and networking of accident data. The aim of this review paper is a description and possible uses

  6. STATE OF CALIFORNIA -GENERAL SERVICES -RISK AND INSURANCE MANAGEMENT STATE DRIVER ACCIDENT REVIEW

    E-Print Network [OSTI]

    Ponce, V. Miguel

    STATE OF CALIFORNIA - GENERAL SERVICES - RISK AND INSURANCE MANAGEMENT STATE DRIVER ACCIDENT REVIEW STD. 274 (REV. 1/2003) PLEASE PRINT OR TYPE SUPERVISOR'S REVIEW - FOR DEPARTMENTAL ACCIDENT PREVENTION PURPOSE: To have supervisor investigate each driver accident, report facts and circumstances, confirm

  7. Manuscript to appear in Environment, Systems and Decisions CALCULATING NUCLEAR ACCIDENT PROBABILITIES

    E-Print Network [OSTI]

    Boyer, Edmond

    Manuscript to appear in Environment, Systems and Decisions CALCULATING NUCLEAR ACCIDENT there is no authoritative, comprehensive and public historical record of nuclear power plant accidents, we reconstructed a nuclear accident dataset from peer-reviewed and other literature. We found that, in a sample of five

  8. Chapter 13 Employee Health and Safety 13.01 Safety Policy and Accident Reporting

    E-Print Network [OSTI]

    Sheridan, Jennifer

    Chapter 13 Employee Health and Safety 13.01 Safety Policy and Accident Reporting General Safety. If an accident occurs, this responsibility includes making an adequate investigation and taking necessary are responsible for following established safety procedures and using protective equipment. Safety and Accident

  9. BP Texas City accident: weak signais or sheer power? Jean Christophe Le Coze,

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    BP Texas City accident: weak signais or sheer power? Jean Christophe Le Coze, Research the interprétation of the BP Texas City accident. While bringing a lot of empirical data on a wide range for sensitising the data. It however clearly indicated a référence to the Columbia Accident Investigation Board

  10. Reflection on a model of accident reporting to help to implement efficient prevention strategies

    E-Print Network [OSTI]

    Boyer, Edmond

    Reflection on a model of accident reporting to help to implement efficient prevention strategies to deliver the relevant action plan especially to control occupational accidents. The aim of the article is to present our approach to analyze the classical Heinrich's model of occupational accidents and the classical

  11. Aircraft Accident Prevention: Loss-of-Control Analysis Harry G. Kwatny

    E-Print Network [OSTI]

    Kwatny, Harry G.

    Aircraft Accident Prevention: Loss-of-Control Analysis Harry G. Kwatny , Jean-Etienne T. Dongmo NASA Langley Research Center, MS 161, Hampton, VA, 23681. The majority of fatal aircraft accidents that during the ten year period 1997-2006, 59% of fatal aircraft accidents were associated with Loss

  12. Radionuclide release calculations for selected severe accident scenarios

    SciTech Connect (OSTI)

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. (Battelle Columbus Div., OH (USA))

    1990-08-01T23:59:59.000Z

    This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs.

  13. Viewing device for electron-beam equipment

    SciTech Connect (OSTI)

    Nasyrov, R.S.

    1985-06-01T23:59:59.000Z

    Viewing devices are used to observe melting, welding, and so on in vacuum systems, an it is necessary to protect the windows from droplets and vapor. A viewing device for electron-beam equipment is described in which the viewing tube and mounting flange are made as a tubular ball joint enclosed in a steel bellows, which render the viewing device flexible. Bending the viewing tube in the intervals between observations protects the viewing window from sputtering and from drops of molten metal.

  14. Accident Reporting Policy Outline the policy regarding accident reporting on University of Michigan (U-M) vehicles.

    E-Print Network [OSTI]

    Kirschner, Denise

    owned by U-M are covered by the U-M self insurance program administered by Risk Management. Policy 1. An accident is defined as any incident that causes damage to persons or property. 2. In the glove box of every

  15. Type B Accident Investigation Board Report of the January 20, 1998, Electrical Accident at the Casa Grande Substation,South of Phoenix, Arizona

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type-B Accident Investigation Board appointed by Michael S.Cowan, Chief Program Officer, Western Area Power Administration.

  16. Localized lesions induced by sup 137 Cs during the Goiania accident

    SciTech Connect (OSTI)

    Oliveira, A.R.; Brandao-Mello, C.E.; Valverde, N.J.; Farina, R.; Curado, M.P. (Industrias Nucleares do Brasil S.A., Rio de Janeiro (Brazil))

    1991-01-01T23:59:59.000Z

    A description is given of initial symptoms and clinical observations regarding acute localized radiation lesions in 28 persons exposed to 137Cs during the Goiania radiological accident. Specialized procedures to estimate the extent and gravity of the lesions and establish a therapeutic strategy, as well as to anticipate the prognosis in each case, are briefly discussed. Measures taken for reduction of pain and inflammation are noted, and an explanation is given for difficulties encountered due to adverse working conditions and the serious clinical manifestations presented by various patients concomitantly with their lesions. Also noted is the difficulty in obtaining credible information regarding exposure, such as source-to-object distance, duration of exposure, and source activity, which precluded dosimetry studies in most cases.

  17. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect (OSTI)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01T23:59:59.000Z

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

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

    SciTech Connect (OSTI)

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

    2003-01-01T23:59:59.000Z

    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.

  19. Wide field of view telescope

    DOE Patents [OSTI]

    Ackermann, Mark R. (Albuquerque, NM); McGraw, John T. (Placitas, NM); Zimmer, Peter C. (Albuquerque, NM)

    2008-01-15T23:59:59.000Z

    A wide field of view telescope having two concave and two convex reflective surfaces, each with an aspheric surface contour, has a flat focal plane array. Each of the primary, secondary, tertiary, and quaternary reflective surfaces are rotationally symmetric about the optical axis. The combination of the reflective surfaces results in a wide field of view in the range of approximately 3.8.degree. to approximately 6.5.degree.. The length of the telescope along the optical axis is approximately equal to or less than the diameter of the largest of the reflective surfaces.

  20. Accident at Three Mile Island: the human dimensions

    SciTech Connect (OSTI)

    Sills, D.L.; Wolf, C.P.; Shelanski, V.B. (eds.)

    1982-01-01T23:59:59.000Z

    A separate abstract was prepared for each of the 19 chapters, divided according to the following Parts: (1) Public Perceptions of Nuclear Energy; (2) Local Responses to Nuclear Plants; (3) Institutional Responsibilities for Nuclear Energy; (4) The Interaction of Social and Technical Systems; and (5) Implications for Public Policy. All of the abstracts will appear in Energy Abstracts for Policy Analysis (EAPA); three will appear in Energy Research Abstracts (ERA). At the request of the President's Commission on the Accident at Three Mile Island (the Kemeny Commission), the Social Science Research Council commissioned social scientists to write a series of papers on the human dimensions of the event. This volume includes those papers, in revised and expanded form, and a comprehensive bibliography of published and unpublished social science research on the accident and its aftermath.