Sample records for leaks damage accidents

  1. Type B Accident Investigation of the Mineral Oil Leak Discovered on January 8, 2001, Resulting in Property Damage at the Atlas Facility, Los Alamos National Laboratory

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Acting Chief Operating Officer for Defense Programs, Ralph E. Erickson.

  2. 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.

  3. 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.

  4. 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.

  5. Assessment of light water reactor fuel damage during a reactivity initiated accident

    SciTech Connect (OSTI)

    MacDonald, P.E.; Seiffert, S.L.; Martinson, Z.R.; McCardell, R.K.; Owen, D.E.; Fukuda, S.K.

    1980-01-01T23:59:59.000Z

    This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO/sub 2/. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies.

  6. Leak detection/verification

    SciTech Connect (OSTI)

    Krhounek, V.; Zdarek, J.; Pecinka, L. [Nuclear Research Institute, Rez (Czech Republic)

    1997-04-01T23:59:59.000Z

    Loss of coolant accident (LOCA) experiments performed as part of a Leak Before Break (LBB) analysis are very briefly summarized. The aim of these experiments was to postulate the leak rates of the coolant. Through-wall cracks were introduced into pipes by fatigue cycling and hydraulically loaded in a test device. Measurements included coolant pressure and temperature, quantity of leaked coolant, displacement of a specimen, and acoustic emission. Small cracks were plugged with particles in the coolant during testing. It is believed that plugging will have no effect in cracks with leak rates above 35 liters per minute. The leak rate safety margin of 10 is sufficient for cracks in which the leak rate is more than 5 liters per minute.

  7. Precursors to potential severe core damage accidents: 1997 -- A status report. Volume 26

    SciTech Connect (OSTI)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-11-01T23:59:59.000Z

    This report describes the five operational events in 1997 that affected five commercial light-water reactors (LWRs) and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by first computer-screening the 1997 licensee event reports from commercial LWRs to identify those events that could be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1996 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  8. Best Management Practice #3: Distribution System Audits, Leak...

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

    Leaks in distribution systems are caused by a number of factors, including pipe corrosion, high system pressure, construction disturbances, frost damage, damaged joints, and...

  9. Precursors to potential severe core damage accidents. A status report, 1982--1983

    SciTech Connect (OSTI)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W. [and others] [and others

    1997-04-01T23:59:59.000Z

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  10. Precursors to potential severe core damage accidents: 1995 A status report

    SciTech Connect (OSTI)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A. [and others] and others

    1997-04-01T23:59:59.000Z

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  11. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    SciTech Connect (OSTI)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States); [Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01T23:59:59.000Z

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

  12. 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

  13. Reducing Your Leak Rate Without Repairing Leaks 

    E-Print Network [OSTI]

    Beals, C.

    2005-01-01T23:59:59.000Z

    As plant personnel know, repairing compressed air leaks can be an expensive, labor intensive and never-ending process. This article discusses ways plant personnel can reduce and maintain their leak rate at a lower level ...

  14. Waste transfer leaks technical basis document

    SciTech Connect (OSTI)

    ZIMMERMAN, B.D.

    2003-03-22T23:59:59.000Z

    This document provides technical support for the onsite radiological and toxicological, and offsite toxicological, portions of the waste transfer leak accident presented in the Documented Safety Analysis. It provides the technical basis for frequency and consequence bin selection, and selection of safety SSCs and TSRs.

  15. Design and fabrication of a maneuverable robot for in-pipe leak detection

    E-Print Network [OSTI]

    Wu, You, S.M. Massachusetts Institute of Technology

    2014-01-01T23:59:59.000Z

    Leaks in pipelines have been causing a significant amount of financial losses and serious damages to the community and the environment. The recent development of in-pipe leak detection technologies at Massachusetts Institute ...

  16. Reducing Your Leak Rate Without Repairing Leaks

    E-Print Network [OSTI]

    Beals, C.

    2005-01-01T23:59:59.000Z

    . It discusses how pressure/flow controllers, variable speed and variable displacement compressors, automation, and addressing critical plant pressures allow plant personnel to lower the header pressure, which eliminates artificial demand and controls the leak...

  17. Natural Gas Pipeline Leaks Across Washington, DC Robert B. Jackson,,,

    E-Print Network [OSTI]

    Jackson, Robert B.

    Natural Gas Pipeline Leaks Across Washington, DC Robert B. Jackson,,, * Adrian Down, Nathan G increased in recent decades, but incidents involving natural gas pipelines still cause an average of 17 fatalities and $133 M in property damage annually. Natural gas leaks are also the largest anthropogenic

  18. Precursors to potential severe core damage accidents: 1992, A status report. Volume 17, Main report and Appendix A

    SciTech Connect (OSTI)

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Jansen, J.M.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States); Lau, W.; Salyer, W.D. [Reliability and Performance Associates (United States)

    1993-12-01T23:59:59.000Z

    Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 {times} 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process.

  19. 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.

  20. Precursors to potential severe core damage accidents: 1994, a status report. Volume 21: Main report and appendices A--H

    SciTech Connect (OSTI)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01T23:59:59.000Z

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

  1. 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’

  2. Leak detection aid

    DOE Patents [OSTI]

    Steeper, Timothy J. (Graniteville, SC)

    1989-01-01T23:59:59.000Z

    A leak detection apparatus and method for detecting leaks across an O-ring sealing a flanged surface to a mating surface is an improvement in a flanged surface comprising a shallow groove following O-ring in communication with an entrance and exit port intersecting the shallow groove for injecting and withdrawing, respectively, a leak detection fluid, such as helium. A small quantity of helium injected into the entrance port will flow to the shallow groove, past the O-ring and to the exit port.

  3. Leak detection aid

    DOE Patents [OSTI]

    Steeper, T.J.

    1989-12-26T23:59:59.000Z

    A leak detection apparatus and method for detecting leaks across an O-ring sealing a flanged surface to a mating surface is an improvement in a flanged surface comprising a shallow groove following O-ring in communication with an entrance and exit port intersecting the shallow groove for injecting and withdrawing, respectively, a leak detection fluid, such as helium. A small quantity of helium injected into the entrance port will flow to the shallow groove, past the O-ring and to the exit port. 2 figs.

  4. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01T23:59:59.000Z

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

  5. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices E (Sections E.1--E.8). Volume 2, Part 3A

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01T23:59:59.000Z

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful.

  6. Gaseous leak detector

    DOE Patents [OSTI]

    Juravic, Jr., Frank E. (Aurora, IL)

    1988-01-01T23:59:59.000Z

    In a short path length mass-spectrometer type of helium leak detector wherein the helium trace gas is ionized, accelerated and deflected onto a particle counter, an arrangement is provided for converting the detector to neon leak detection. The magnetic field of the deflection system is lowered so as to bring the non linear fringe area of the magnetic field across the ion path, thereby increasing the amount of deflection of the heavier neon ions.

  7. Improved gaseous leak detector

    DOE Patents [OSTI]

    Juravic, F.E. Jr.

    1983-10-06T23:59:59.000Z

    In a short path length mass-spectrometer type of helium leak detector wherein the helium trace gas is ionized, accelerated and deflected onto a particle counter, an arrangement is provided for converting the detector to neon leak detection. The magnetic field of the deflection system is lowered so as to bring the nonlinear fringe area of the magnetic field across the ion path, thereby increasing the amount of deflection of the heavier neon ions.

  8. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

    1994-06-01T23:59:59.000Z

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

  9. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States)] [and others

    1994-06-01T23:59:59.000Z

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

  10. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendices F-H, Volume 2, Part 4

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

    1994-06-01T23:59:59.000Z

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

  11. Sensitive hydrogen leak detector

    DOE Patents [OSTI]

    Myneni, Ganapati Rao (Yorktown, VA)

    1999-01-01T23:59:59.000Z

    A sensitive hydrogen leak detector system using passivation of a stainless steel vacuum chamber for low hydrogen outgassing, a high compression ratio vacuum system, a getter operating at 77.5 K and a residual gas analyzer as a quantitative hydrogen sensor.

  12. Hazardous fluid leak detector

    DOE Patents [OSTI]

    Gray, Harold E. (Las Vegas, NV); McLaurin, Felder M. (Las Vegas, NV); Ortiz, Monico (Las Vegas, NV); Huth, William A. (Las Vegas, NV)

    1996-01-01T23:59:59.000Z

    A device or system for monitoring for the presence of leaks from a hazardous fluid is disclosed which uses two electrodes immersed in deionized water. A gas is passed through an enclosed space in which a hazardous fluid is contained. Any fumes, vapors, etc. escaping from the containment of the hazardous fluid in the enclosed space are entrained in the gas passing through the enclosed space and transported to a closed vessel containing deionized water and two electrodes partially immersed in the deionized water. The electrodes are connected in series with a power source and a signal, whereby when a sufficient number of ions enter the water from the gas being bubbled through it (indicative of a leak), the water will begin to conduct, thereby allowing current to flow through the water from one electrode to the other electrode to complete the circuit and activate the signal.

  13. Natural gas leak mapper

    DOE Patents [OSTI]

    Reichardt, Thomas A. (Livermore, CA); Luong, Amy Khai (Dublin, CA); Kulp, Thomas J. (Livermore, CA); Devdas, Sanjay (Albany, CA)

    2008-05-20T23:59:59.000Z

    A system is described that is suitable for use in determining the location of leaks of gases having a background concentration. The system is a point-wise backscatter absorption gas measurement system that measures absorption and distance to each point of an image. The absorption measurement provides an indication of the total amount of a gas of interest, and the distance provides an estimate of the background concentration of gas. The distance is measured from the time-of-flight of laser pulse that is generated along with the absorption measurement light. The measurements are formated into an image of the presence of gas in excess of the background. Alternatively, an image of the scene is superimosed on the image of the gas to aid in locating leaks. By further modeling excess gas as a plume having a known concentration profile, the present system provides an estimate of the maximum concentration of the gas of interest.

  14. Aspects of leak detection

    SciTech Connect (OSTI)

    Chivers, T.C. [Berkeley Technology Centre, Glos (United Kingdom)

    1997-04-01T23:59:59.000Z

    A requirement of a Leak before Break safety case is that the leakage from the through wall crack be detected prior to any growth leading to unacceptable failure. This paper sets out to review some recent developments in this field. It does not set out to be a comprehensive guide to all of the methods available. The discussion concentrates on acoustic emission and how the techniques can be qualified and deployed on operational plant.

  15. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage

    SciTech Connect (OSTI)

    Budnitz, R.J. [Future Resources Associates, Inc., Berkeley, CA (United States); Davis, P.R. [PRD Consulting (United States); Ravindra, M.K.; Tong, W.H. [EQE International, Inc., Irvine, CA (United States)

    1994-08-01T23:59:59.000Z

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ``internal initiators.`` This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10{sup {minus}7}/year.

  16. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1. Volume 5: Analysis of core damage frequency from seismic events during mid-loop operations

    SciTech Connect (OSTI)

    Budnitz, R.J. [Future Resources Associates, Inc., Berkeley, CA (United States); Davis, P.R. [PRD Consulting (United States); Ravindra, M.K.; Tong, W.H. [EQE International, Inc., Irvine, CA (United States)

    1994-08-01T23:59:59.000Z

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1) and the other at Sandia National Laboratories studying a boiling water reactor (Grand Gulf). Both the Brookhaven and Sandia projects have examined only accidents initiated by internal plant faults--so-called ``internal initiators.`` This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Surry Unit 1. All of the many systems modeling assumptions, component non-seismic failure rates, and human error rates that were used in the internal-initiator study at Surry have been adopted here, so that the results of the two studies can be as comparable as possible. Both the Brookhaven study and this study examine only two shutdown plant operating states (POSs) during refueling outages at Surry, called POS 6 and POS 10, which represent mid-loop operation before and after refueling, respectively. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POSs 6 and 10. The results of the analysis are that the core-damage frequency of earthquake-initiated accidents during refueling outages in POS 6 and POS 10 is found to be low in absolute terms, less than 10{sup {minus}6}/year.

  17. Leak test fitting

    DOE Patents [OSTI]

    Pickett, Patrick T. (Kettering, OH)

    1981-01-01T23:59:59.000Z

    A hollow fitting for use in gas spectrometry leak testing of conduit joints is divided into two generally symmetrical halves along the axis of the conduit. A clip may quickly and easily fasten and unfasten the halves around the conduit joint under test. Each end of the fitting is sealable with a yieldable material, such as a piece of foam rubber. An orifice is provided in a wall of the fitting for the insertion or detection of helium during testing. One half of the fitting also may be employed to test joints mounted against a surface.

  18. 225-B Pool Cell 5 Liner Leak Investigation

    SciTech Connect (OSTI)

    Rasmussen, J.H., Westinghouse Hanford

    1996-06-07T23:59:59.000Z

    This document describes the actions taken to confirm and respond to a very small (0.046 ml/min) leak in the stainless steel liner of Hanford`s Waste Encapsulation and Storage Facility (WESF) storage pool cell 5 in Building 225-B. Manual level measurements confirmed a consistent weekly accumulation of 0.46 liters of water in the leak detection grid sump below the pool cell 5 liner. Video inspections and samples point to the capsule storage pool as the source of the water. The present leak rate corresponds to a decrease of only 0.002 inches per week in the pool cell water level, and consequently does not threaten any catastrophic loss of pool cell shielding and cooling water. The configuration of the pool cell liner, sump system, and associated risers will limit the short-term consequences of even a total liner breach to a loss of 1 inch in pool cell level. The small amount of demineralized pool cell water which has been in contact with the concrete structure is not enough to cause significant structural damage. However, ongoing water-concrete interaction increases. The pool cell leak detection sump instrumentation will be modified to improve monitoring of the leak rate in the future. Weekly manual sump level measurements continue in the interim. Contingency plans are in place to relocate the pool cell 5 capsules if the leak worsens.

  19. Leaking Pipelines: Doctoral Student Family Formation

    E-Print Network [OSTI]

    Serrano, Christyna M.

    2008-01-01T23:59:59.000Z

    Sari M. “Why the Academic Pipeline Leaks: Fewer Men thanone reason the academic pipeline leaks. 31 Blair-Loy, Mary.to leak out of the “academic pipeline. ” The term “academic

  20. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-09-01T23:59:59.000Z

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  1. Precursors to potential severe core damage accidents: 1992, a status report; Volume 18: Appendices B, C, D, E, F, and G

    SciTech Connect (OSTI)

    NONE

    1993-12-01T23:59:59.000Z

    This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; {open_quote}interesting{close_quote} events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports.

  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. Analysis of SX farm leak histories -- Historical leak model (HLM)

    SciTech Connect (OSTI)

    Fredenburg, E.A.

    1998-08-20T23:59:59.000Z

    This report uses readily available historical information to better define the volume, chemical composition, and Cs-137/Sr-90 amounts for leaks that have occurred in the past for tanks SX-108, SX-109, SX-111, and SX-112. In particular a Historical Leak Model (HLM) is developed that is a month by month reconciliation of tank levels, fill records, and calculated boil-off rates for these tanks. The HLM analysis is an independent leak estimate that reconstructs the tank thermal histories thereby deriving each tank`s evaporative volume loss and by difference, its unaccounted losses as well. The HLM analysis was meant to demonstrate the viability of its approach, not necessarily to establish the HLM leak estimates as being definitive. Past leak estimates for these tanks have invariably resorted to soil wetting arguments but the extent of soil contaminated by each leak has always been highly uncertain. There is also a great deal of uncertainty with the HLM that was not quantified in this report, but will be addressed later. These four tanks (among others) were used from 1956 to 1975 for storage of high-level waste from the Redox process at Hanford. During their operation, tank waste temperatures were often as high as 150 C (300 F), but were more typically around 130 C. The primary tank cooling was by evaporation of tank waste and therefore periodic replacement of lost volume with water was necessary to maintain each tank`s inventory. This active reflux of waste resulted in very substantial turnovers in tank inventory as well as significant structural degradation of these tanks. As a result of the loss of structural integrity, each of these tanks leaked during their active periods of operation. Unfortunately, the large turnover in tank volume associated with their reflux cooling has made a determination of leak volumes very difficult. During much of these tanks operational histories, inventory losses because of evaporative cooling could have effectively masked any volume loss due to leak. However, careful comparison with reported tank levels during certain periods clearly show unaccounted volume losses for many tanks. As a result of the HLM analysis, SX-108, SX-109, SX-111, and SX-112 all show clear evidence of unaccounted volume losses during the period 1958 to 1975. Likewise, the HLM does not show similar unaccounted volume losses for tank SX-105, a tank with no reported leak history, verifying that the HLM is consistent with SX-105 not leaking. These unaccounted volume losses establish the leak start date and rate, and when propagated over time show that SX-108 lost 203 kgal followed by SX-109 at 111. SX-111 at 55, and SX-112 at 44 kgal.0664 These leak volumes represent maximum or upper bounds estimates of each leak and are in total volume about six times the previous leak estimates.

  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. 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,...

  7. 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,...

  8. Analysis of risk reduction methods for interfacing system LOCAs (loss-of-coolant accidents) at PWRs

    SciTech Connect (OSTI)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1988-01-01T23:59:59.000Z

    The Reactor Safety Study (WASH-1400) predicted that Interfacing System Loss-of-Coolant Accidents (ISL) events were significant contributors to risk even though they were calculated to be relatively low frequency events. However, there are substantial uncertainties involved in determining the probability and consequences of the ISL sequences. For example, the assumed valve failure modes, common cause contributions and the location of the break/leak are all uncertain and can significantly influence the predicted risk from ISL events. In order to provide more realistic estimates for the core damage frequencies (CDFs) and a reduction in the magnitude of the uncertainties, a reexamination of ISL scenarios at PWRs has been performed by Brookhaven National Laboratory. The objective of this study was to investigate the vulnerability of pressurized water reactor designs to ISLs and identify any improvements that could significantly reduce the frequency/risk of these events.

  9. Detecting Air Leaks | Department of Energy

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

    See if you can rattle them, since movement means possible air leaks. If you can see daylight around a door or window frame, then the door or window leaks. You can usually seal...

  10. Stochastic Consequence Analysis for Waste Leaks

    SciTech Connect (OSTI)

    HEY, B.E.

    2000-05-31T23:59:59.000Z

    This analysis evaluates the radiological consequences of potential Hanford Tank Farm waste transfer leaks. These include ex-tank leaks into structures, underneath the soil, and exposed to the atmosphere. It also includes potential misroutes, tank overflow

  11. 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

  12. 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

  13. 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

  14. Leak detection capability in CANDU reactors

    SciTech Connect (OSTI)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01T23:59:59.000Z

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  15. Vacuum leak detector and method

    DOE Patents [OSTI]

    Edwards, Jr., David (7 Brown's La., Bellport, NY 11713)

    1983-01-01T23:59:59.000Z

    Apparatus and method for detecting leakage in a vacuum system involves a moisture trap chamber connected to the vacuum system and to a pressure gauge. Moisture in the trap chamber is captured by freezing or by a moisture adsorbent to reduce the residual water vapor pressure therein to a negligible amount. The pressure gauge is then read to determine whether the vacuum system is leaky. By directing a stream of carbon dioxide or helium at potentially leaky parts of the vacuum system, the apparatus can be used with supplemental means to locate leaks.

  16. Leak detection on an ethylene pipeline

    SciTech Connect (OSTI)

    Hamande, A.; Condacse, V.; Modisette, J.

    1995-12-31T23:59:59.000Z

    A model-based leak detection system has been in operation on the Solvay et Cie ethylene pipeline from Antwerp to Jemeppe on Sambre since 1989. The leak detection system, which is the commercial product PLDS of Modisette Associations, Inc., was originally installed by the supplier. Since 1991, all system maintenance and configuration changes have been done by Solvay et Cie personnel. Many leak tests have been performed, and adjustments have been made in the configuration and the automatic tuning parameters. The leak detection system is currently able to detect leaks of 2 tonnes/hour in 11 minutes with accurate location. Larger leaks are detected in about 2 minutes. Leaks between 0.5 and 1 tonne per hour are detected after several hours. (The nominal mass flow in the pipeline is 15 tonnes/hour, with large fluctuations.) Leaks smaller than 0.5 tonnes per hour are not detected, with the alarm thresholds set at levels to avoid false alarms. The major inaccuracies of the leak detection system appear to be associated with the ethylene temperatures.

  17. High sensitivity leak detection method and apparatus

    DOE Patents [OSTI]

    Myneni, G.R.

    1994-09-06T23:59:59.000Z

    An improved leak detection method is provided that utilizes the cyclic adsorption and desorption of accumulated helium on a non-porous metallic surface. The method provides reliable leak detection at superfluid helium temperatures. The zero drift that is associated with residual gas analyzers in common leak detectors is virtually eliminated by utilizing a time integration technique. The sensitivity of the apparatus of this disclosure is capable of detecting leaks as small as 1 [times] 10[sup [minus]18] atm cc sec[sup [minus]1]. 2 figs.

  18. High sensitivity leak detection method and apparatus

    DOE Patents [OSTI]

    Myneni, Ganapatic R. (Grafton, VA)

    1994-01-01T23:59:59.000Z

    An improved leak detection method is provided that utilizes the cyclic adsorption and desorption of accumulated helium on a non-porous metallic surface. The method provides reliable leak detection at superfluid helium temperatures. The zero drift that is associated with residual gas analyzers in common leak detectors is virtually eliminated by utilizing a time integration technique. The sensitivity of the apparatus of this disclosure is capable of detecting leaks as small as 1.times.10.sup.-18 atm cc sec.sup.-1.

  19. Saving Money with Air and Gas Leak Surveys

    E-Print Network [OSTI]

    Woodruff, D.

    2010-01-01T23:59:59.000Z

    uncorrected air leaks and gas leaks cost your businesses time and money as well as being environmentally unfriendly. ? Air Leak Surveys ? Nitrogen Leak Surveys ? Gas Leak Survey (H2, O2, Natural Gas) ? Steam Leak Surveys ? Steam Trap Surveys ? Safe... sites per year ? Member of ISNetworld, and Browz. ? Security Checks o Petro Chemical Energy employee background checks performed by DISA ? Drugs & Alcohol Free Workplace o Petro Chemical Energy employees are tested for Drugs and Alcohol prior...

  20. Leak checker data logging system

    DOE Patents [OSTI]

    Gannon, Jeffrey C. (Arlington, TX); Payne, John J. (Waterman, IL)

    1996-01-01T23:59:59.000Z

    A portable, high speed, computer-based data logging system for field testing systems or components located some distance apart employs a plurality of spaced mass spectrometers and is particularly adapted for monitoring the vacuum integrity of a long string of a superconducting magnets such as used in high energy particle accelerators. The system provides precise tracking of a gas such as helium through the magnet string when the helium is released into the vacuum by monitoring the spaced mass spectrometers allowing for control, display and storage of various parameters involved with leak detection and localization. A system user can observe the flow of helium through the magnet string on a real-time basis hour the exact moment of opening of the helium input valve. Graph reading can be normalized to compensate for magnet sections that deplete vacuum faster than other sections between testing to permit repetitive testing of vacuum integrity in reduced time.

  1. Leak checker data logging system

    DOE Patents [OSTI]

    Gannon, J.C.; Payne, J.J.

    1996-09-03T23:59:59.000Z

    A portable, high speed, computer-based data logging system for field testing systems or components located some distance apart employs a plurality of spaced mass spectrometers and is particularly adapted for monitoring the vacuum integrity of a long string of a superconducting magnets such as used in high energy particle accelerators. The system provides precise tracking of a gas such as helium through the magnet string when the helium is released into the vacuum by monitoring the spaced mass spectrometers allowing for control, display and storage of various parameters involved with leak detection and localization. A system user can observe the flow of helium through the magnet string on a real-time basis hour the exact moment of opening of the helium input valve. Graph reading can be normalized to compensate for magnet sections that deplete vacuum faster than other sections between testing to permit repetitive testing of vacuum integrity in reduced time. 18 figs.

  2. Modular, High-Volume Fuel Cell Leak-Test Suite and Process

    SciTech Connect (OSTI)

    Ru Chen; Ian Kaye

    2012-03-12T23:59:59.000Z

    Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

  3. Rapid communication Mapping urban pipeline leaks: Methane leaks across Boston

    E-Print Network [OSTI]

    Jackson, Robert B.

    transmission and distribution pipelines for natural gas in the U. S. cause an average of 17 fatalities, 68 signatures w20& lighter (m ¼ À57.8&, Æ1.6& s.e., n ¼ 8). Repairing leaky natural gas distribution systems injuries, and $133 M in property damage each year (PHMSA, 2012). A natural gas pipeline explosion in San

  4. 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)

  5. 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

  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. Adversaries and Information Leaks Geoffrey Smith

    E-Print Network [OSTI]

    Smith, Geoffrey

    Adversaries and Information Leaks (Tutorial) Geoffrey Smith School of Computing and Information-Verlag Berlin Heidelberg 2008 #12;384 G. Smith ­ The program c has direct access to the sensitive information

  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

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

  9. 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

  10. 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

  11. SINGLE-SHELL TANKS LEAK INTEGRITY ELEMENTS/SX FARM LEAK CAUSES AND LOCATIONS - 12127

    SciTech Connect (OSTI)

    VENETZ TJ; WASHENFELDER D; JOHNSON J; GIRARDOT C

    2012-01-25T23:59:59.000Z

    Washington River Protection Solutions, LLC (WRPS) developed an enhanced single-shell tank (SST) integrity project in 2009. An expert panel on SST integrity was created to provide recommendations supporting the development of the project. One primary recommendation was to expand the leak assessment reports (substitute report or LD-1) to include leak causes and locations. The recommendation has been included in the M-045-9IF Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) as one of four targets relating to SST leak integrity. The 241-SX Farm (SX Farm) tanks with leak losses were addressed on an individual tank basis as part of LD-1. Currently, 8 out of 23 SSTs that have been reported to having a liner leak are located in SX Farm. This percentage was the highest compared to other tank farms which is why SX Farm was analyzed first. The SX Farm is comprised of fifteen SSTs built 1953-1954. The tanks are arranged in rows of three tanks each, forming a cascade. Each of the SX Farm tanks has a nominal I-million-gal storage capacity. Of the fifteen tanks in SX Farm, an assessment reported leak losses for the following tanks: 241-SX-107, 241-SX-108, 241-SX-109, 241-SX-111, 241-SX-112, 241-SX-113, 241-SX-114 and 241-SX-115. The method used to identify leak location consisted of reviewing in-tank and ex-tank leak detection information. This provided the basic data identifying where and when the first leaks were detected. In-tank leak detection consisted of liquid level measurement that can be augmented with photographs which can provide an indication of the vertical leak location on the sidewall. Ex-tank leak detection for the leaking tanks consisted of soil radiation data from laterals and drywells near the tank. The in-tank and ex-tank leak detection can provide an indication of the possible leak location radially around and under the tank. Potential leak causes were determined using in-tank and ex-tank information that is not directly related to leak detection. In-tank parameters can include temperature of the supernatant and sludge, types of waste, and chemical determination by either transfer or sample analysis. Ex-tank information can be assembled from many sources including design media, construction conditions, technical specifications, and other sources. Five conditions may have contributed to SX Farm tank liner failure including: tank design, thermal shock, chemistry-corrosion, liner behavior (bulging), and construction temperature. Tank design did not apparently change from tank to tank for the SX Farm tanks; however, there could be many unknown variables present in the quality of materials and quality of construction. Several significant SX Farm tank design changes occurred from previous successful tank farm designs. Tank construction occurred in winter under cold conditions which could have affected the ductile to brittle transition temperature of the tanks. The SX Farm tanks received high temperature boiling waste from REDOX which challenged the tank design with rapid heat up and high temperatures. All eight of the leaking SX Farm tanks had relatively high rate of temperature rise. Supernatant removal with subsequent nitrate leaching was conducted in all but three of the eight leaking tanks prior to leaks being detected. It is possible that no one characteristic of the SX Farm tanks could in isolation from the others have resulted in failure. However, the application of so many stressors - heat up rate, high temperature, loss of corrosion protection, and tank design - working jointly or serially resulted in their failure. Thermal shock coupled with the tank design, construction conditions, and nitrate leaching seem to be the overriding factors that can lead to tank liner failure. The distinction between leaking and sound SX Farm tanks seems to center on the waste types, thermal conditions, and nitrate leaching.

  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. Analysis of main steam isolation valve leakage in design basis accidents using MELCOR 1.8.6 and RADTRAD.

    SciTech Connect (OSTI)

    Salay, Michael (United States Nuclear Regulatory Commission, Washington, D.C.); Kalinich, Donald A.; Gauntt, Randall O.; Radel, Tracy E.

    2008-10-01T23:59:59.000Z

    Analyses were performed using MELCOR and RADTRAD to investigate main steam isolation valve (MSIV) leakage behavior under design basis accident (DBA) loss-of-coolant (LOCA) conditions that are presumed to have led to a significant core melt accident. Dose to the control room, site boundary and LPZ are examined using both approaches described in current regulatory guidelines as well as analyses based on best estimate source term and system response. At issue is the current practice of using containment airborne aerosol concentrations as a surrogate for the in-vessel aerosol concentration that exists in the near vicinity of the MSIVs. This study finds current practice using the AST-based containment aerosol concentrations for assessing MSIV leakage is non-conservative and conceptually in error. A methodology is proposed that scales the containment aerosol concentration to the expected vessel concentration in order to preserve the simplified use of the AST in assessing containment performance under assumed DBA conditions. This correction is required during the first two hours of the accident while the gap and early in-vessel source terms are present. It is general practice to assume that at {approx}2hrs, recovery actions to reflood the core will have been successful and that further core damage can be avoided. The analyses performed in this study determine that, after two hours, assuming vessel reflooding has taken place, the containment aerosol concentration can then conservatively be used as the effective source to the leaking MSIV's. Recommendations are provided concerning typical aerosol removal coefficients that can be used in the RADTRAD code to predict source attenuation in the steam lines, and on robust methods of predicting MSIV leakage flows based on measured MSIV leakage performance.

  15. acoustic leak detection: Topics by E-print Network

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

    PLoS ONE 9(10): e Lawrence, Rick L. 8 Design and fabrication of a maneuverable robot for in-pipe leak detection MIT - DSpace Summary: Leaks in pipelines have been causing...

  16. New system pinpoints leaks in ethylene pipeline

    SciTech Connect (OSTI)

    Hamande, A. [Solvay et Cie, Jemeppe sur Sambre (Belgium); Condacse, V.; Modisette, J. [Modisette Associates, Inc., Houston, TX (United States)

    1995-04-01T23:59:59.000Z

    A model-based leak detection, PLDS, developed by Modisette Associates, Inc., Houston has been operating on the Solvay et Cie ethylene pipeline since 1989. The 6-in. pipeline extends from Antwerp to Jemeppe sur Sambre, a distance of 73.5 miles and is buried at a depth of 3 ft. with no insulation. Except for outlets to flares, located every 6 miles for test purposes, there are no injections or deliveries along the pipeline. Also, there are block valves, which are normally open, at each flare location. This paper reviews the design and testing procedures used to determine the system performance. These tests showed that the leak system was fully operational and no false alarms were caused by abrupt changes in inlet/outlet flows of the pipeline. It was confirmed that leaks larger than 2 tonnes/hr. (40 bbl/hr) are quickly detected and accurately located. Also, maximum leak detection sensitivity is 1 tonne/hr. (20 bbl/hr) with a detection time of one hour. Significant operational, configuration, and programming issues also were found during the testing program. Data showed that temperature simulations needed re-examining for improvement since accurate temperature measurements are important. This is especially true for ethylene since its density depends largely on temperature. Another finding showed the averaging period of 4 hrs. was too long and a 1 to 2 hr. interval was better.

  17. Managing an Effective Leak Sealing Program

    E-Print Network [OSTI]

    Rinz, W. H.

    1980-01-01T23:59:59.000Z

    An on-line leak sealing program is an extremely effective method of cost savings to industrial plants. The dollars a plant saves can be direct and dramatic as in an avoided system shut-down or subtle and analytical as in a long term maintenance...

  18. Experiences with leak rate calculations methods for LBB application

    SciTech Connect (OSTI)

    Grebner, H.; Kastner, W.; Hoefler, A.; Maussner, G. [and others

    1997-04-01T23:59:59.000Z

    In this paper, three leak rate computer programs for the application of leak before break analysis are described and compared. The programs are compared to each other and to results of an HDR Reactor experiment and two real crack cases. The programs analyzed are PIPELEAK, FLORA, and PICEP. Generally, the different leak rate models are in agreement. To obtain reasonable agreement between measured and calculated leak rates, it was necessary to also use data from detailed crack investigations.

  19. Double Shell Tank AY-102 Radioactive Waste Leak Investigation

    SciTech Connect (OSTI)

    Washenfelder, Dennis J.

    2014-04-10T23:59:59.000Z

    PowerPoint. The objectives of this presentation are to: Describe Effort to Determine Whether Tank AY-102 Leaked; Review Probable Causes of the Tank AY-102 Leak; and, Discuss Influence of Leak on Hanford’s Double-Shell Tank Integrity Program.

  20. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal floods during mid-loop operations. Volume 4

    SciTech Connect (OSTI)

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

    1994-07-01T23:59:59.000Z

    The major objective of the Surry internal flood analysis was to provide an improved understanding of the core damage scenarios arising from internal flood-related events. The mean core damage frequency of the Surry plant due to internal flood events during mid-loop operations is 4.8E-06 per year, and the 5th and 95th percentiles are 2.2E-07 and 1.8E-05 per year, respectively. Some limited sensitivity calculations were performed on three plant improvement options. The most significant result involves modifications of intake-level structure on the canal, which reduced core damage frequency contribution from floods in mid-loop by about 75%.

  1. 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,...

  2. Accident Analysis and Prevention 59 (2013) 8793 Contents lists available at SciVerse ScienceDirect

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    2013-01-01T23:59:59.000Z

    Accident Analysis and Prevention 59 (2013) 87­93 Contents lists available at SciVerse ScienceDirect Accident Analysis and Prevention journal homepage: www.elsevier.com/locate/aap Analysis of U.S. freight t Derailments are the most common type of freight-train accidents in the United States. Derailments cause damage

  3. Method for mapping a natural gas leak

    DOE Patents [OSTI]

    Reichardt, Thomas A. (Livermore, CA); Luong, Amy Khai (Dublin, CA); Kulp, Thomas J. (Livermore, CA); Devdas, Sanjay (Albany, CA)

    2009-02-03T23:59:59.000Z

    A system is described that is suitable for use in determining the location of leaks of gases having a background concentration. The system is a point-wise backscatter absorption gas measurement system that measures absorption and distance to each point of an image. The absorption measurement provides an indication of the total amount of a gas of interest, and the distance provides an estimate of the background concentration of gas. The distance is measured from the time-of-flight of laser pulse that is generated along with the absorption measurement light. The measurements are formatted into an image of the presence of gas in excess of the background. Alternatively, an image of the scene is superimposed on the image of the gas to aid in locating leaks. By further modeling excess gas as a plume having a known concentration profile, the present system provides an estimate of the maximum concentration of the gas of interest.

  4. 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

  5. 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.

  6. 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...

  7. 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...

  8. 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.

  9. LEAK: A source term generator for evaluating release rates from leaking vessels

    SciTech Connect (OSTI)

    Clinton, J.H.

    1994-09-01T23:59:59.000Z

    An interactive computer code for estimating the rate of release of any one of several materials from a leaking tank or broken pipe leading from a tank is presented. It is generally assumed that the material in the tank is liquid. Materials included in the data base are acetonitrile, ammonia, carbon tetrachloride, chlorine, chlorine trifluoride, fluorine, hydrogen fluoride, nitric acid, nitrogen tetroxide, sodium hydroxide, sulfur hexafluoride, sulfuric acid, and uranium hexafluoride. Materials that exist only as liquid and/or vapor over expected ranges of temperature and pressure can easily be added to the data base file. The Fortran source code for LEAK and the data file are included with this report.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. One-Piece Leak-Proof Battery

    DOE Patents [OSTI]

    Verhoog, Roelof (Bordeaux, FR)

    1999-03-23T23:59:59.000Z

    The casing of a leak-proof one-piece battery is made of a material comprising a mixture of at least a matrix based on polypropylene and an alloy of a polyamide and a polypropylene. The ratio of the matrix to the alloy is in the range 0.5 to 6 by weight. The alloy forms elongate arborescent inclusions in the matrix such that, on average, the largest dimension of a segment of the arborescence is at least twenty times the smallest dimension of the segment.

  15. ANNUAL MAINTENANCE AND LEAK TESTING FOR THE 9975 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Trapp, D.

    2014-08-25T23:59:59.000Z

    The purpose of this document is to provide step-by-step instructions for the annual helium leak test certification and maintenance of the 9975 Shipping Package.

  16. Margins in high temperature leak-before-break assessments

    SciTech Connect (OSTI)

    Budden, P.J.; Hooton, D.G.

    1997-04-01T23:59:59.000Z

    Developments in the defect assessment procedure R6 to include high-temperature mechanisms in Leak-before-Break arguments are described. In particular, the effect of creep on the time available to detect a leak and on the crack opening area, and hence leak rate, is discussed. The competing influence of these two effects is emphasized by an example. The application to Leak-before-Break of the time-dependent failure assessment diagram approach for high temperature defect assessment is then outlined. The approach is shown to be of use in assessing the erosion of margins by creep.

  17. 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.

  18. Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

    2012-09-10T23:59:59.000Z

    Measurements of several radionuclides within environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Company website following the recent tsunami-initiated catastrophe were evaluated for the purpose of identifying the source term, reconstructing the release mechanisms, and estimating the extent of the release. 136Cs/137Cs and 134Cs/137Cs ratios identified Units 1-3 as the major source of radioactive contamination to the surface soil close to the facility. A trend was observed between the fraction of the total core inventory released for a number of fission product isotopes and their corresponding Gibbs Free Energy of formation for the primary oxide form of the isotope, suggesting that release was dictated primarily by chemical volatility driven by temperature and reduction potential within the primary containment vessels of the vented reactors. The absence of any major fractionation beyond volatilization suggested all coolant had evaporated by the time of venting. High estimates for the fraction of the total inventory released of more volatile species (Te, Cs, I) indicated the damage to fuel bundles was likely extensive, minimizing any potential containment due to physical migration of these species through the fuel matrix and across the cladding wall. 238Pu/239,240Pu ratios close-in and at 30 km from the facility indicated that the damaged reactors were the major contributor of Pu to surface soil at the source but that this contribution likely decreased rapidly with distance from the facility. The fraction of the total Pu inventory released to the environment from venting units 1 and 3 was estimated to be ~0.003% based upon Pu/Cs isotope ratios relative to the within-reactor modeled inventory prior to venting and was consistent with an independent model evaluation that considered chemical volatility based upon measured fission product release trends. Significant volatile radionuclides within the spent fuel at the time of venting but not as yet observed and reported within environmental samples are suggested as potential analytes of concern for future environmental surveys around the site.

  19. A new blowdown compensation scheme for boiler leak detection

    E-Print Network [OSTI]

    Marquez, Horacio J.

    considers the blowdown effect in industrial boiler operation. This adds to the efficiency of recent advancesA new blowdown compensation scheme for boiler leak detection A. M. Pertew ,1 X. Sun ,1 R. Kent in identification-based leak detection techniques of boiler steam- water systems. Keywords: Industrial Boilers, Tube

  20. 241-AY-102 Leak Detection Pit Drain Line Inspection Report

    SciTech Connect (OSTI)

    Boomer, Kayle D. [Washington River Protection Solutions, LLC (United States); Engeman, Jason K. [Washington River Protection Solutions, LLC (United States); Gunter, Jason R. [Washington River Protection Solutions, LLC (United States); Joslyn, Cameron C. [Washington River Protection Solutions, LLC (United States); Vazquez, Brandon J. [Washington River Protection Solutions, LLC (United States); Venetz, Theodore J. [Washington River Protection Solutions, LLC (United States); Garfield, John S. [AEM Consulting (United States)

    2014-01-20T23:59:59.000Z

    This document provides a description of the design components, operational approach, and results from the Tank AY-102 leak detection pit drain piping visual inspection. To perform this inspection a custom robotic crawler with a deployment device was designed, built, and operated by IHI Southwest Technologies, Inc. for WRPS to inspect the 6-inch leak detection pit drain line.

  1. Heat exchanger with leak detecting double wall tubes

    SciTech Connect (OSTI)

    Bieberbach, George (Tampa, FL); Bongaards, Donald J. (Seminole, FL); Lohmeier, Alfred (Tampa, FL); Duke, James M. (St. Petersburg, all of, FL)

    1981-01-01T23:59:59.000Z

    A straight shell and tube heat exchanger utilizing double wall tubes and three tubesheets to ensure separation of the primary and secondary fluid and reliable leak detection of a leak in either the primary or the secondary fluids to further ensure that there is no mixing of the two fluids.

  2. 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...

  3. 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,

  4. 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

  5. Proceedings of the seminar on leak before break in reactor piping and vessels

    SciTech Connect (OSTI)

    Faidy, C. [ed.] [Electricite de France, Villeurbanne (France); Gilles, P. [ed.] [Framatome, Paris (France)

    1997-04-01T23:59:59.000Z

    The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  6. Leak-Path Factor Analysis for the Nuclear Materials Storage Facility

    SciTech Connect (OSTI)

    Shaffer, C.; Leonard, M.

    1999-06-13T23:59:59.000Z

    Leak-path factors (LPFs) were calculated for the Nuclear Materials Storage Facility (NMSF) located in the Plutonium Facility, Building 41 at the Los Alamos National Laboratory Technical Area 55. In the unlikely event of an accidental fire powerful enough to fail a container holding actinides, the subsequent release of oxides, modeled as PuO{sub 2} aerosols, from the facility and into the surrounding environment was predicted. A 1-h nondestructive assay (NDA) laboratory fire accident was simulated with the MELCOR severe accident analysis code. Fire-driven air movement along with wind-driven air infiltration transported a portion of these actinides from the building. This fraction is referred to as the leak-path factor. The potential effect of smoke aerosol on the transport of the actinides was investigated to verify the validity of neglecting the smoke as conservative. The input model for the NMSF consisted of a system of control volumes, flow pathways, and surfaces sufficient to model the thermal-hydraulic conditions within the facility and the aerosol transport data necessary to simulate the transport of PuO{sub 2} particles. The thermal-hydraulic, heat-transfer, and aerosol-transport models are solved simultaneously with data being exchanged between models. A MELCOR input model was designed such that it would reproduce the salient features of the fire per the corresponding CFAST calculation. Air infiltration into and out of the facility would be affected strongly by wind-driven differential pressures across the building. Therefore, differential pressures were applied to each side of the building according to guidance found in the ASHRAE handbook using a standard-velocity head equation with a leading multiplier to account for the orientation of the wind with the building. The model for the transport of aerosols considered all applicable transport processes, but the deposition within the building clearly was dominated by gravitational settling.

  7. Pressure Change Measurement Leak Testing Errors

    SciTech Connect (OSTI)

    Pryor, Jeff M [ORNL] [ORNL; Walker, William C [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    A pressure change test is a common leak testing method used in construction and Non-Destructive Examination (NDE). The test is known as being a fast, simple, and easy to apply evaluation method. While this method may be fairly quick to conduct and require simple instrumentation, the engineering behind this type of test is more complex than is apparent on the surface. This paper intends to discuss some of the more common errors made during the application of a pressure change test and give the test engineer insight into how to correctly compensate for these factors. The principals discussed here apply to ideal gases such as air or other monoatomic or diatomic gasses; however these same principals can be applied to polyatomic gasses or liquid flow rate with altered formula specific to those types of tests using the same methodology.

  8. 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.

  9. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01T23:59:59.000Z

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  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. 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

  13. NCI-Frederick Safety and Environmental Compliance Manual 03/2013 B-2. Accident Reporting

    E-Print Network [OSTI]

    Wlodawer, Alexander

    and Environmental Compliance Manual 03/2013 B-2-2 Occupational injury - Is identified as any bodily damageNCI-Frederick Safety and Environmental Compliance Manual 03/2013 B-2-1 B-2. Accident Reporting I or reasonably could result in injury, illness, or property damage. Reporting is mandatory in order that: 1

  14. 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.

  15. Statistical approaches to leak detection for geological sequestration

    E-Print Network [OSTI]

    Haidari, Arman S

    2011-01-01T23:59:59.000Z

    Geological sequestration has been proposed as a way to remove CO? from the atmosphere by injecting it into deep saline aquifers. Detecting leaks to the atmosphere will be important for ensuring safety and effectiveness of ...

  16. Design of a Novel In-Pipe Reliable Leak Detector

    E-Print Network [OSTI]

    Chatzigeorgiou, Dimitris

    Leakage is the major factor for unaccounted losses in every pipe network around the world (oil, gas, or water). In most cases, the deleterious effects associated with the occurrence of leaks may present serious economical ...

  17. allowable leak rates: Topics by E-print Network

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

    A; Provot, N 2011-01-01 36 Journal of Bioenergetics and Biomembranes, Vol. 31, No. 5, 1999 Mitochondrial Proton Leak and the Uncoupling Proteins Biology and Medicine Websites...

  18. The feasibility of electrophoretic repair of impoundment leaks

    E-Print Network [OSTI]

    Han, Ji-Seok

    2002-01-01T23:59:59.000Z

    finding, repairing and testing the leaks, are tedious, expensive, and dangerous to workers. Electrophoretic repair technique is an innovative, economic, and safe method to repair the leakage of impoundments. A suspension of clay particles is induced...

  19. Robot design for leak detection in water-pipe systems

    E-Print Network [OSTI]

    Choi, Changrak

    2012-01-01T23:59:59.000Z

    Leaks are major problem that occur in the water pipelines all around the world. Several reports indicate loss of around 20 to 30 percent of water in the distribution of water through water pipe systems. Such loss of water ...

  20. 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.

  1. Property Loss / Damage Report Damage Loss Details

    E-Print Network [OSTI]

    Ponce, V. Miguel

    Property Loss / Damage Report Damage Loss Details Date & Time of Damage / Loss: Type of damage / loss: Location - specific address / room: Project / Grant associated with damage / loss - grant Police: When was damage / loss first discovered - BY WHOM: Pictures available or attached? Was personal

  2. 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.

  3. 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,...

  4. 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

  5. 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

  6. Damaged Fuel Experiment DF-1

    SciTech Connect (OSTI)

    Gasser, R.D.; Fryer, C.P.; Gauntt, R.O.; Marshall, A.C.; Reil, K.O.; Stalker, K.T.

    1990-01-01T23:59:59.000Z

    A series of in-pile experiments addressing LWR severe fuel damage phenomena has been conducted in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The ACRR Debris Formation and Relocation (DF) experiments are quasi-separate effects tests that provide a data base for the development and verification of models for LWR severe core damage accidents. The first experiment in this series, DF-1, was performed on March 15, 1984, and the results are presented in this report. The DF-1 experiment examined the effects of low initial clad oxidation conditions on fuel damage and relocation processes. The DF-1 test assembly consisted of a nine-rod square-matrix bundle that employed PWR-type fuel rods with a 0.5-m fissile length. The fuel rods were composed of 10% enriched UO{sub 2} pellets within a zircaloy-4 cladding. Steam flowed through the test bundle at flow rates varying between 0.5 and 3 g/s, and the ACRR maintained a peak power level of 1.5 MW during the high temperature oxidation phase of the test inducing {approximately}8.5 kW fission power and {approximately}20 kW peak oxidation power in the assembly. Visual observation showed early clad relocation and partial blockage formation at the grid spacer location accompanied by production of a dense aerosol. Posttest cross sections show liquefaction losses of fuel in excess of 10 volume percent, as well as large fractional losses of cladding material from the upper two-thirds of the bundle. The quantity of hydrogen measured during the test was consistent with the observed magnitude of cladding oxidation. Oxidation driven heating rates of 25 K/s and peak temperatures in excess of 2525 K were observed. The analyses, interpretation, and application of these results to severe fuel damage accidents are discussed. 27 refs., 118 figs., 23 tabs.

  7. 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.

  8. Oil/gas collector/separator for underwater oil leaks

    DOE Patents [OSTI]

    Henning, Carl D. (Livermore, CA)

    1993-01-01T23:59:59.000Z

    An oil/gas collector/separator for recovery of oil leaking, for example, from an offshore or underwater oil well. The separator is floated over the point of the leak and tethered in place so as to receive oil/gas floating, or forced under pressure, toward the water surface from either a broken or leaking oil well casing, line, or sunken ship. The separator is provided with a downwardly extending skirt to contain the oil/gas which floats or is forced upward into a dome wherein the gas is separated from the oil/water, with the gas being flared (burned) at the top of the dome, and the oil is separated from water and pumped to a point of use. Since the density of oil is less than that of water it can be easily separated from any water entering the dome.

  9. Apparatus and method for detecting leaks in piping

    DOE Patents [OSTI]

    Trapp, Donald J. (Aiken, SC)

    1994-01-01T23:59:59.000Z

    A method and device for detecting the location of leaks along a wall or piping system, preferably in double-walled piping. The apparatus comprises a sniffer probe, a rigid cord such as a length of tube attached to the probe on one end and extending out of the piping with the other end, a source of pressurized air and a source of helium. The method comprises guiding the sniffer probe into the inner pipe to its distal end, purging the inner pipe with pressurized air, filling the annulus defined between the inner and outer pipe with helium, and then detecting the presence of helium within the inner pipe with the probe as is pulled back through the inner pipe. The length of the tube at the point where a leak is detected determines the location of the leak in the pipe.

  10. Apparatus and method for detecting leaks in piping

    DOE Patents [OSTI]

    Trapp, D.J.

    1994-12-27T23:59:59.000Z

    A method and device are disclosed for detecting the location of leaks along a wall or piping system, preferably in double-walled piping. The apparatus comprises a sniffer probe, a rigid cord such as a length of tube attached to the probe on one end and extending out of the piping with the other end, a source of pressurized air and a source of helium. The method comprises guiding the sniffer probe into the inner pipe to its distal end, purging the inner pipe with pressurized air, filling the annulus defined between the inner and outer pipe with helium, and then detecting the presence of helium within the inner pipe with the probe as is pulled back through the inner pipe. The length of the tube at the point where a leak is detected determines the location of the leak in the pipe. 2 figures.

  11. RPP-ENV-39658 Revision 0 Hanford SX-Farm Leak Assessments Report

    E-Print Network [OSTI]

    M. E. Johnson; J. G. Field; Revision Rpp-env

    2010-01-01T23:59:59.000Z

    U.S. Department of Energy developed a process to reassess selected tank leak estimates (volumes and inventories), and to update single-shell tank leak and unplanned release volumes and inventory estimates as emergent field data is obtained (RPP-32681, Process to Assess Tank Farm Leaks in Support of Retrieval and Closure Planning). This process does not represent a formal tank leak assessment in accordance with procedure TFC-ENG-CHEM-D-42, “Tank Leak Assessment Process. ” This report documents reassessment of past leaks in the 241-SX Tank Farm. Tank waste loss events were reassessed for tanks 241-SX-104, 241-SX-107, 241-SX-108,

  12. Leak before break application in French PWR plants under operation

    SciTech Connect (OSTI)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01T23:59:59.000Z

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  13. 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

  14. 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

  15. Drill-in fluids control formation damage

    SciTech Connect (OSTI)

    Halliday, W.S. (Baker Hughes Inteq, Houston, TX (United States))

    1994-12-01T23:59:59.000Z

    Several factors led to development, oil company interest in, and use of payzone drilling fluids, including operator concern about maximizing well production, increasing acceptance of horizontal drilling and openhole completion popularity. This article discusses water-base drill-in'' fluid systems and applications. Payzone damage, including fine solids migration, clay swelling and solids invasion, reduces effective formation permeability, which results in lower production rates. Formation damage is often caused by invasion of normal drilling fluids that contain barite or bentonite. Drill-in systems are designed with special bridging agents to minimize invasion. Several bridging materials designed to form effective filter cake for instantaneous leak-off control can be used. Bridging materials are also designed to minimize stages and time required to clean up wells before production. Fluids with easy-to-remove bridging agents reduce completion costs. Drill-in fluid bridging particles can often be removed more thoroughly than those in standard fluids.

  16. Improved geomembrane damage/leak detection through co-extrusion technology

    SciTech Connect (OSTI)

    Messmer, D.P.; Cadwallader, M. (Gundle Lining Systems, Inc., Houston, TX (United States))

    1994-04-01T23:59:59.000Z

    There has been a considerable advancement in technology available for providing a barrier system in the containment and storage of waste materials. Natural soil liners several feet in thickness have been augmented by factory-produced, synthetic materials that have permeability coefficients several orders of magnitude lower than any natural soil system. To carry the systems approach one step farther, engineers use multiple layers of synthetics separated at times by layers of clay offering a redundant composite barrier to protect the groundwater. Each geosynthetic material offers its own unique contribution to the system based upon its physical characteristics. Co-extrusion -- the process of combining two or more materials into a single product, through a single process -- is now revolutionizing the liner industry.

  17. Analysis and design of an in-pipe system for water leak detection

    E-Print Network [OSTI]

    Chatzigeorgiou, Dimitris M

    2010-01-01T23:59:59.000Z

    Leaks are a major factor for unaccounted water losses in almost every water distribution network. Pipeline leak may result, for example, from bad workmanship or from any destructive cause, due to sudden changes of pressure, ...

  18. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    SciTech Connect (OSTI)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01T23:59:59.000Z

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  19. Harms of Unintentional Leaks during Volume Targeted Pressure Support Ventilation

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Harms of Unintentional Leaks during Volume Targeted Pressure Support Ventilation Sonia Khirani1 Background: Volume targeted pressure support ventilation (VT-PSV) is a hybrid mode increasingly used. The objective of the study was to determine the ability of home ventilators to maintain the preset minimal VT

  20. Mineral formation during simulated leaks of Hanford waste tanks

    E-Print Network [OSTI]

    Flury, Markus

    Mineral formation during simulated leaks of Hanford waste tanks Youjun Deng a , James B. Harsh a at the US DOE Hanford Site, Washington, caus- ing mineral dissolution and re-precipitation upon contact with subsurface sediments. The main mineral precipitation and transformation pathways were studied in solutions

  1. AIR SEALING Seal air leaks and save energy!

    E-Print Network [OSTI]

    Oak Ridge National Laboratory

    AIR SEALING Seal air leaks and save energy! W H A T I S A I R L E A K A G E ? Ventilation is fresh air that enters a house in a controlled manner to exhaust excess moisture and reduce odors and stuffiness. Air leakage, or infiltration, is outside air that enters a house uncontrollably through cracks

  2. T Plant secondary containment and leak detection upgrades

    SciTech Connect (OSTI)

    Carlson, T.A.

    1995-10-19T23:59:59.000Z

    The W-259 project will provide upgrades to the 2706-T/TA Facility to comply with Federal and State of Washington environmental regulations for secondary containment and leak detection. The project provides decontamination activities supporting the environmental restoration mission and waste management operations on the Hanford Site.

  3. Methodology to quantify leaks in aerosol sampling system components

    E-Print Network [OSTI]

    Vijayaraghavan, Vishnu Karthik

    2004-11-15T23:59:59.000Z

    and that approach was used to measure the sealing integrity of a CAM and two kinds of filter holders. The methodology involves use of sulfur hexafluoride as a tracer gas with the device being tested operated under dynamic flow conditions. The leak rates...

  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. 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)

  6. Hanford Single-Shell Tank Leak Causes and Locations - 241-B Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L. [Washington River Protection Systems, Richland, WA (United States); Harlow, Donald G. [Washington River Protection Systems, Richland, WA (United States)

    2013-07-11T23:59:59.000Z

    This document identifies 241-B Tank Farm (B Farm) leak cause and locations for the 100 series leaking tank (241-B-107) identified in RPP-RPT-49089, Hanford B-Farm Leak Inventory Assessments Report. This document satisfies the B Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  7. Mathematical Properties of Pump-Leak Models of Cell Volume Control and Electrolyte Balance

    E-Print Network [OSTI]

    Weinberger, Hans

    Mathematical Properties of Pump-Leak Models of Cell Volume Control and Electrolyte Balance Yoichiro using pump-leak models, a system of differential algebraic equations that de- scribes the balance and stability of steady states for a general class of pump-leak models. We treat two cases. When the ion channel

  8. Mitochondrial proton leak and the uncoupling protein 1 homologues J.A. Stuart aYb

    E-Print Network [OSTI]

    Stuart, Jeffrey A.

    Review Mitochondrial proton leak and the uncoupling protein 1 homologues J.A. Stuart aYb , S 2000 Abstract Mitochondrial proton leak is the largest single contributor to the standard metabolic rate (SMR) of a rat, accounting for about 20% of SMR. Yet the mechanisms by which proton leak occurs

  9. Quantitative studies of severe fuel damage using delayed neutron data

    SciTech Connect (OSTI)

    Bauer, T.H.; Braid, T.H. (Argonne National Lab., IL (USA)); Schleisiek, K. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1990-01-01T23:59:59.000Z

    A technique to quantify fuel damage in an LMR through analysis of delayed neutron data is presented, which is suitable for analysis of both small-scale in-pile experiments and full-scale plants. Validating analyses are described for five in-pile severe accident simulations performed within the SLSF and Mol 7C test programs. Comparison is made of measured and calculated amounts of fuel damage. 8 refs., 5 figs., 2 tabs.

  10. Quantitative studies of severe fuel damage using delayed neutron data

    SciTech Connect (OSTI)

    Bauer, T.H.; Braiel, T.H. (Argonne National Lab., IL (United States)); Schleisiek, K. (Kernforschungszentrum Karlsruhe GmbH (Germany))

    1992-09-01T23:59:59.000Z

    In this paper, a technique is presented to quantify fuel damage in a liquid-metal reactor through fast-running computer analysis of delayed neutron data, suitable for analysis of both small-scale in-pile experiments and full-scale plants. Validating analyses are described for five in-pile severe accident simulations performed within the Sodium Loop Safety Facility and Mol-7C test programs. Comparison is made of measured and calculated amounts of fuel damage.

  11. 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

  12. 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

  13. 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

  14. 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...

  15. Simulations of argon accident scenarios in the ATLAS experimental cavern a safety analysis

    E-Print Network [OSTI]

    Balda, F

    2002-01-01T23:59:59.000Z

    Some characteristic accidents in the ATLAS experimental cavern (UX15) are simulated by means of STAR-CD, a code using the "Finite-Volume" method. These accidents involve different liquid argon leaks from the barrel cryostat of the detector, thus causing the dispersion of the argon into the Muon Chamber region and the evaporation of the liquid. The subsequent temperature gradients and distribution of argon concentrations, as well as their evolution in time are simulated and discussed, with the purpose of analysing the dangers related to asphyxiation and to contact with cryogenic fluids for the working personnel. A summary of the theory that stands behind the code is also given. In order to validate the models, an experimental test on a liquid argon spill performed earlier is simulated, showing that the program is able to output reliable results. At the end, some safety-related recommendations are listed.

  16. Preliminary analysis of graphite dust releasing behavior in accident for HTR

    SciTech Connect (OSTI)

    Peng, W.; Yang, X. Y.; Yu, S. Y.; Wang, J. [Inst. of Nuclear and New Energy Technology, Tsinghua Univ., Beijing100084 (China)

    2012-07-01T23:59:59.000Z

    The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. This study investigated the flow of graphite dust in helium mainstream. The analysis of the stresses acting on the graphite dust indicated that gas drag played the absolute leading role. Based on the understanding of the importance of gas drag, an experimental system is set up for the research of dust releasing behavior in accident. Air driven by centrifugal fan is used as the working fluid instead of helium because helium is expensive, easy to leak which make it difficult to seal. The graphite particles, with the size distribution same as in HTR, are added to the experiment loop. The graphite dust releasing behavior at the loss-of-coolant accident will be investigated by a sonic nozzle. (authors)

  17. 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.

  18. 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.

  19. MCO combustible gas management leak test acceptance criteria

    SciTech Connect (OSTI)

    SHERRELL, D.L.

    1999-05-11T23:59:59.000Z

    Existing leak test acceptance criteria for mechanically sealed and weld sealed multi-canister overpacks (MCO) were evaluated to ensure that MCOs can be handled and stored in stagnant air without compromising the Spent Nuclear Fuel Project's overall strategy to prevent accumulation of combustible gas mixtures within MCO's or within their surroundings. The document concludes that the integrated leak test acceptance criteria for mechanically sealed and weld sealed MCOs (1 x 10{sup -5} std cc/sec and 1 x 10{sup -7} std cc/sec, respectively) are adequate to meet all current and foreseeable needs of the project, including capability to demonstrate compliance with the NFPA 60 Paragraph 3-3 requirement to maintain hydrogen concentrations [within the air atmosphere CSB tubes] t or below 1 vol% (i.e., at or below 25% of the LFL).

  20. TOWARDS SHARING OF DATA FROM ACCIDENTS WITH CHEMICALS J.P. Pineau, J.F. Lechaudel

    E-Print Network [OSTI]

    Boyer, Edmond

    in Flixborough (explosion with 29 fatalities) in 1974, Sandoz Bäle (fire with release of chemicals in water substance such äs a major emission, fire or explosion leading to serious damage to human health, transportation and use of chemicals were at the origin of very severe accidents. The manufacture of explosives

  1. Accident, Illness and Liability Coverage Risk Management in the 4-H Youth Development Program

    E-Print Network [OSTI]

    Tullos, Desiree

    1 Accident, Illness and Liability Coverage Risk Management in the 4-H Youth Development Program for injuries or damages to person or property of others (tort liability), when the volunteer is: · Working volunteer, and enrolled or listed with the Oregon State University Extension 4-H Youth Development Program

  2. Leak detection, monitoring, and mitigation technology trade study update

    SciTech Connect (OSTI)

    HERTZEL, J.S.

    1998-11-10T23:59:59.000Z

    This document is a revision and update to the initial report that describes various leak detection, monitoring, and mitigation (LDMM) technologies that can be used to support the retrieval of waste from the single-shell tanks (SST) at the Hanford Site. This revision focuses on the improvements in the technical performance of previously identified and useful technologies, and it introduces new technologies that might prove to be useful.

  3. Environmental Stigma Damages: Speculative Damages in Environmental Tort Cases

    E-Print Network [OSTI]

    Johnson, E. Jean

    1997-01-01T23:59:59.000Z

    contami- nation causing environmental damage cannot be seen,Damages: Speculative Damages in Environmental Tort Cases E.in cases of environmental damage, primar- ily because it is

  4. Electrical detection of liquid lithium leaks from pipe joints

    SciTech Connect (OSTI)

    Schwartz, J. A., E-mail: jschwart@pppl.gov; Jaworski, M. A.; Mehl, J.; Kaita, R.; Mozulay, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

    2014-11-15T23:59:59.000Z

    A test stand for flowing liquid lithium is under construction at Princeton Plasma Physics Laboratory. As liquid lithium reacts with atmospheric gases and water, an electrical interlock system for detecting leaks and safely shutting down the apparatus has been constructed. A defense in depth strategy is taken to minimize the risk and impact of potential leaks. Each demountable joint is diagnosed with a cylindrical copper shell electrically isolated from the loop. By monitoring the electrical resistance between the pipe and the copper shell, a leak of (conductive) liquid lithium can be detected. Any resistance of less than 2 k? trips a relay, shutting off power to the heaters and pump. The system has been successfully tested with liquid gallium as a surrogate liquid metal. The circuit features an extensible number of channels to allow for future expansion of the loop. To ease diagnosis of faults, the status of each channel is shown with an analog front panel LED, and monitored and logged digitally by LabVIEW.

  5. Electrical detection of liquid lithium leaks from pipe joints

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

    Schwartz, J. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA; Jaworski, M. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA; Mehl, J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA; Mozulay, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA

    2014-11-01T23:59:59.000Z

    A test stand for flowing liquid lithium is under construction at Princeton Plasma Physics Laboratory. As liquid lithium reacts with atmospheric gases and water, an electrical interlock system for detecting leaks and safely shutting down the apparatus has been constructed. A defense in depth strategy is taken to minimize the risk and impact of potential leaks. Each demountable joint is diagnosed with a cylindrical copper shell electrically isolated from the loop. By monitoring the electrical resistance between the pipe and the copper shell, a leak of (conductive) liquid lithium can be detected. Any resistance of less than 2 k#2; trips a relay, shutting off power to the heaters and pump. The system has been successfully tested with liquid gallium as a surrogate liquid metal. The circuit features an extensible number of channels to allow for future expansion of the loop. To ease diagnosis of faults, the status of each channel is shown with an analog front panel LED, and monitored and logged digitally by LabVIEW.

  6. 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.

  7. 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

  8. 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...

  9. 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...

  10. 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

  11. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 2, Part 1C: Analysis of core damage frequency from internal events for plant operational State 5 during a refueling outage, Main report (Sections 11--14)

    SciTech Connect (OSTI)

    Whitehead, D. [Sandia National Labs., Albuquerque, NM (United States); Darby, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Yakle, J. [Science Applications International Corp., Albuquerque, NM (United States)] [and others

    1994-06-01T23:59:59.000Z

    This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power analyses performed on Grand Gulf.

  12. Hurricane-damaged Gulf of Mexico pipeline repaired with cold forging

    SciTech Connect (OSTI)

    Lewis, G. (Texaco Pipeline Inc., Houma, LA (United States)); DeGruy, P. (Texaco Inc., New Orleans, LA (United States)); Avery, L. (Big Inch Marine Systems Inc., Lafayette, LA (United States))

    1993-05-03T23:59:59.000Z

    Damage to Texaco Pipeline Inc.'s Eugene Island Pipeline System (EIPS) in last year's Hurricane Andrew prompted a complex repair project unique for the Gulf of Mexico. Damage, suffered when the anchor of a runaway semisubmersible drilling rig crashed into the 20-in. EPIS during the height of the storm, caused the pipeline to fail under pressure within 48 hr. after start-up following the storm. The paper describes the importance of the EIPS; system safety; Andrew's damage; locating the leak; repair options; the chosen system; mechanical bonding; end connectors and ball flanges; and diving operations.

  13. 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...

  14. 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

  15. 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

  16. 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...

  17. 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

  18. 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

  19. 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

  20. Hanford Single-Shell Tank Leak Causes and Locations - 241-U Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-12-02T23:59:59.000Z

    This document identifies 241-U Tank Farm (U Farm) leak causes and locations for the 100 series leaking tanks (241-U-104, 241-U-110, and 241-U-112) identified in RPP-RPT-50097, Rev. 0, Hanford 241-U Farm Leak Inventory Assessment Report. This document satisfies the U-Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  1. Hanford Single Shell Tank Leak Causes and Locations - 241-TX Farm

    SciTech Connect (OSTI)

    Girardot, C. L.; Harlow, D> G.

    2014-07-22T23:59:59.000Z

    This document identifies 241-TX Tank Farm (TX Farm) leak causes and locations for the 100 series leaking tanks (241-TX-107 and 241-TX-114) identified in RPP-RPT-50870, Rev. 0, Hanford 241-TX Farm Leak Inventory Assessment Report. This document satisfies the TX Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  2. Hanford Single-Shell Tank Leak Causes and Locations - 241-C Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-07-30T23:59:59.000Z

    This document identifies 241-C Tank Farm (C Farm) leak causes and locations for the 100 series leaking tanks (241-C-101 and 241-C-105) identified in RPP-RPT-33418, Rev. 2, Hanford C-Farm Leak Inventory Assessments Report. This document satisfies the C Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  3. Hanford Single-Shell Tank Leak Causes and Locations - 241-T Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2014-05-15T23:59:59.000Z

    This document identifies 241-T Tank Farm (T Farm) leak causes and locations for the 100 series leaking tanks (241-T-106 and 241-T-111) identified in RPP-RPT-55084, Rev. 0, Hanford 241-T Farm Leak Inventory Assessment Report. This document satisfies the T Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  4. Long-wave infrared imaging of vegetation for detecting leaking CO2 gas

    E-Print Network [OSTI]

    Shaw, Joseph A.

    Long-wave infrared imaging of vegetation for detecting leaking CO2 gas Jennifer E. Johnson Joseph A for detecting leaking CO2 gas Jennifer E. Johnson,a Joseph A. Shaw,a Rick Lawrence,b Paul W. Nugent,a Laura M of these calibrated imagers is imaging of vegetation for CO2 gas leak detection. During a four-week period

  5. Beyond Leaks: Demand-side Strategies for Improving Compressed Air Efficiency

    E-Print Network [OSTI]

    Howe, B.; Scales, B.

    Beyond Leaks: Demand-side Strategies for Bill Howe, PE Director, Corporate Energy Services E Source, Inc. Boulder, Colorado SUMMARY Staggering amounts of compressed air are wasted or misapplied in otherwise well run manufacturing...-maintained plants lose about 10 percent of compressed air to leaks, while many more lose over 50 percent. In addition to leaks, wasteful application of compressed air can eat up another 5 to 40 percent of compressed air volume-even in otherwise well...

  6. Hanford Single-Shell Tank Leak Causes and Locations - 241-A Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-09-10T23:59:59.000Z

    This document identifies 241-A Tank Farm (A Farm) leak causes and locations for the 100 series leaking tanks (241-A-104 and 241-A-105) identified in RPP-ENV-37956, Hanford A and AX Farm Leak Assessment Report. This document satisfies the A Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  7. A mathematical model for air brake systems in the presence of leaks

    E-Print Network [OSTI]

    Ramaratham, Srivatsan

    2008-10-10T23:59:59.000Z

    of the pneumatic subsystem. . . . . . . . . . . . . . . . . . 20 16 Pressure transients at 722 kPa (90 psi) supply pressure with no leak. 22 17 Schematic of the setup for leak corroboration tests. . . . . . . . . . . 27 18 Comparison of measured and predicted mass... of detecting and locating leaks[6]. Most of the performance tests and visual based inspection tests of the air brake system indirectly correlate pressure in the brake chamber with the torque output, brake pad temperature, push rod strokes etc[7], [8]. More...

  8. Leake County, Mississippi: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal Pwer Plant Jump to:Landowners and Wind EnergyIndiana: Energy Resources JumpPrataHill,LeadingLeake

  9. Air Leaks in Unexpected Places | Department of Energy

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed off Energy.gov. Are you sure you want toworldPower 2010 1A Potential PathAddingAhorreLeaks in

  10. Hydrogen Leak Detection - Low-Cost Distributed Gas Sensors | Department

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking of Blythe SolarContamination Detectorof Energy Leak Detection - Low-Cost

  11. Hanford Single-Shell Tank Leak Causes and Locations - 241-BY and 241-TY Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-11-19T23:59:59.000Z

    This document identifies 241-BY Tank Farm (BY Farm) and 241-TY Tank Farm (TY Farm) leak causes and locations for the 100 series leaking tanks (241-BY-103, 241-TY-103, 241-TY-104, 241-TY-105, and 241-TY-106) identified in RPP-RPT-43704, Hanford BY Farm Leak Assessments Report, and in RPP-RPT-42296, Hanford TY Farm Leak Assessments Report. This document satisfies the BY and TY Farm portion of the target (T04) in Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  12. Ultra high vacuum pumping system and high sensitivity helium leak detector

    DOE Patents [OSTI]

    Myneni, Ganapati Rao (Yorktown, VA)

    1997-01-01T23:59:59.000Z

    An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10.sup.-13 atm cc s.sup.-1. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces backstreaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium.

  13. Ultra high vacuum pumping system and high sensitivity helium leak detector

    DOE Patents [OSTI]

    Myneni, G.R.

    1997-12-30T23:59:59.000Z

    An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10{sup {minus}13} atm cc/s. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces back streaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium. 2 figs.

  14. T-726:Linux-2.6 privilege escalation/denial of service/information leak

    Broader source: Energy.gov [DOE]

    Vulnerabilities have been discovered in the Linux kernel that may lead to a privilege escalation, denial of service or information leak.

  15. KERENA safety concept in the context of the Fukushima accident

    SciTech Connect (OSTI)

    Zacharias, T.; Novotny, C.; Bielor, E. [AREVA NP GmbH, Paul-Gossen-Strasse 100, 91052 Erlangen (Germany)

    2012-07-01T23:59:59.000Z

    Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basic physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)

  16. 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).

  17. 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.

  18. Of Disasters and Dragon Kings: A Statistical Analysis of Nuclear Power Incidents & Accidents

    E-Print Network [OSTI]

    Wheatley, Spencer; Sornette, Didier

    2015-01-01T23:59:59.000Z

    We provide, and perform a risk theoretic statistical analysis of, a dataset that is 75 percent larger than the previous best dataset on nuclear incidents and accidents, comparing three measures of severity: INES (International Nuclear Event Scale), radiation released, and damage dollar losses. The annual rate of nuclear accidents, with size above 20 Million US$, per plant, decreased from the 1950s until dropping significantly after Chernobyl (April, 1986). The rate is now roughly stable at 0.002 to 0.003, i.e., around 1 event per year across the current fleet. The distribution of damage values changed after Three Mile Island (TMI; March, 1979), where moderate damages were suppressed but the tail became very heavy, being described by a Pareto distribution with tail index 0.55. Further, there is a runaway disaster regime, associated with the "dragon-king" phenomenon, amplifying the risk of extreme damage. In fact, the damage of the largest event (Fukushima; March, 2011) is equal to 60 percent of the total damag...

  19. Damage susceptibility tables

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

    conditions, 3) that x-ray induced damage rates for a common material (relatively pure PVC in this example 7,8 ) can be used to normalize different sets of XPS damage...

  20. Intermediate-Scale Laboratory Experiments of Subsurface Flow and Transport Resulting from Tank Leaks

    SciTech Connect (OSTI)

    Oostrom, Martinus; Wietsma, Thomas W.

    2014-09-30T23:59:59.000Z

    Washington River Protection Solutions contracted with Pacific Northwest National Laboratory to conduct laboratory experiments and supporting numerical simulations to improve the understanding of water flow and contaminant transport in the subsurface between waste tanks and ancillary facilities at Waste Management Area C. The work scope included two separate sets of experiments: •Small flow cell experiments to investigate the occurrence of potential unstable fingering resulting from leaks and the limitations of the STOMP (Subsurface Transport Over Multiple Phases) simulator to predict flow patterns and solute transport behavior under these conditions. Unstable infiltration may, under certain conditions, create vertically elongated fingers potentially transporting contaminants rapidly through the unsaturated zone to groundwater. The types of leak that may create deeply penetrating fingers include slow release, long duration leaks in relatively permeable porous media. Such leaks may have occurred below waste tanks at the Hanford Site. •Large flow experiments to investigate the behavior of two types of tank leaks in a simple layered system mimicking the Waste Management Area C. The investigated leaks include a relatively large leak with a short duration from a tank and a long duration leak with a relatively small leakage rate from a cascade line.

  1. Low-cost multispectral vegetation imaging system for detecting leaking CO2 gas

    E-Print Network [OSTI]

    Shaw, Joseph A.

    Low-cost multispectral vegetation imaging system for detecting leaking CO2 gas Justin A. Hogan,1 sequestration sites for possible leaks of the CO2 gas from underground reservoirs, a low-cost multispectral are then flagged for closer inspection with in-situ CO2 sensors. The system is entirely self

  2. Project uses microphones to detect underwater gas leaks Published: 14 Oct 2011

    E-Print Network [OSTI]

    Sóbester, András

    into developing the technology,' said Leighton. Topics: Research and Development, carbon capture use and storage as naturally occurring methane gas leaks. `The current carbon- capture storage facilities have the ability Key Topics: Technology Scientists at Southampton University are employing hydrophones to monitor leaks

  3. System and damage identification of civil structures

    E-Print Network [OSTI]

    Moaveni, Babak

    2007-01-01T23:59:59.000Z

    12 Damage Index Methods. . . . . . . . . . . . . . . .Model Updating for Damage Identification . . . . . . . .298 x Damage Factors and Residual

  4. Internal dissipation and heat leaks in quantum thermodynamic cycles

    E-Print Network [OSTI]

    Luis A. Correa; José P. Palao; Daniel Alonso

    2015-07-06T23:59:59.000Z

    The direction of the steady-state heat currents across a generic quantum system connected to multiple baths may be engineered so as to realize virtually any thermodynamic cycle. In spite of their versatility such continuous energy-conversion systems are generally unable to operate at maximum efficiency due to non-negligible sources of irreversible entropy production. In this paper we introduce a minimal model of irreversible absorption chiller. We identify and characterize the different mechanisms responsible for its irreversibility, namely heat leaks and internal dissipation, and gauge their relative impact in the overall cooling performance. We also propose reservoir engineering techniques to minimize these detrimental effects. Finally, by looking into a known three-qubit embodiment of the absorption cooling cycle, we illustrate how our simple model may help to pinpoint the different sources of irreversibility naturally arising in more complex practical heat devices.

  5. Theory of the leak-rate of seals

    E-Print Network [OSTI]

    B. N. J. Persson; C. Yang

    2008-05-06T23:59:59.000Z

    Seals are extremely useful devices to prevent fluid leakage. However, the exact mechanism of roughness induced leakage is not well understood. We present a theory of the leak-rate of seals, which is based on percolation theory and a recently developed contact mechanics theory. We study both static and dynamics seals. We present molecular dynamics results which show that when two elastic solids with randomly rough surfaces are squeezed together, as a function of increasing magnification or decreasing squeezing pressure, a non-contact channel will percolate when the (relative) projected contact area, A/A_0, is of order 0.4, in accor dance with percolation theory. We suggest a simple experiment which can be used to test the theory.

  6. 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.

  7. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect (OSTI)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18T23:59:59.000Z

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  8. Estimates of the financial consequences of nuclear-power-reactor accidents

    SciTech Connect (OSTI)

    Strip, D.R.

    1982-09-01T23:59:59.000Z

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

  9. 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...

  10. Studies into the Initial Conditions, Flow Rate, and Containment System of Oil Field Leaks in Deep Water

    E-Print Network [OSTI]

    Holder, Rachel

    2013-07-22T23:59:59.000Z

    to contain an oil leak in the field. The dome was found to have satisfactory entrapment in the designed position....

  11. Detection and location of leaks in district heating steam systems: Survey and review of current technology and practices

    SciTech Connect (OSTI)

    Kupperman, D.S.; Raptis, A.C.; Lanham, R.N.

    1992-03-01T23:59:59.000Z

    This report presents the results of a survey undertaken to identify and characterize current practices for detecting and locating leaks in district heating systems, particular steam systems. Currently used technology and practices are reviewed. In addition, the survey was used to gather information that may be important for the application of acoustic leak detection. A few examples of attempts to locate leaks in steam and hot water pipes by correlation of acoustic signals generated by the leaks are also discussed.

  12. 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.

  13. 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.

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. 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

  19. Synthesis of VERCORS and Phebus data in severe accident codes and applications.

    SciTech Connect (OSTI)

    Gauntt, Randall O.

    2010-04-01T23:59:59.000Z

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.

  20. Leak-Tight Welding Experience from the Industrial Assembly of the LHC Cryostats at CERN

    E-Print Network [OSTI]

    Bourcey, N; Chiggiato, P; Limon, P; Mongelluzzo, A; Musso, G; Poncet, A; Parma, V

    2008-01-01T23:59:59.000Z

    The assembly of the approximately 1700 LHC main ring cryostats at CERN involved extensive welding of cryogenic lines and vacuum vessels. More than 6 km of welding requiring leak tightness to a rate better than 1.10-9 mbar.l.s-1 on stainless steel and aluminium piping and envelopes was made, essentially by manual welding but also making use of orbital welding machines. In order to fulfil the safety regulations related to pressure vessels and to comply with the leak-tightness requirements of the vacuum systems of the machine, welds were executed according to high qualification standards and following a severe quality assurance plan. Leak detection by He mass spectrometry was extensively used. Neon leak detection was used successfully to locate leaks in the presence of helium backgrounds. This paper presents the quality assurance strategy adopted for welds and leak detection. It presents the statistics of non-conformities on welds and leaks detected throughout the entire production and the advances in the use...

  1. Leak detection systems for uranium mill tailings impoundments with synthetic liners

    SciTech Connect (OSTI)

    Myers, D.A.; Tyler, S.W.; Gutknecht, P.J.; Mitchell, D.H.

    1983-09-01T23:59:59.000Z

    This study evaluated the performance of existing and alternative leak detection systems for lined uranium mill tailings ponds. Existing systems for detecting leaks at uranium mill tailings ponds investigated in this study included groundwater monitoring wells, subliner drains, and lysimeters. Three alternative systems which demonstrated the ability to locate leaks in bench-scale tests included moisture blocks, soil moisture probes, and a soil resistivity system. Several other systems in a developmental stage are described. For proper performance of leak detection systems (other than groundwater wells and lysimeters), a subgrade is required which assures lateral dispersion of a leak. Methods to enhance dispersion are discussed. Cost estimates were prepared for groundwater monitoring wells, subliner drain systems, and the three experimental systems. Based on the results of this report, it is suggested that groundwater monitoring systems be used as the primary means of leak detection. However, if a more responsive system is required due to site characteristics and groundwater quality criteria, subliner drains are applicable for ponds with uncovered liners. Leak-locating systems for ponds with covered liners require further development. Other recommendations are discussed in the report.

  2. Hanford Single-Shell Tank Leak Causes and Locations - 241-BY and 241-TY Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2014-09-04T23:59:59.000Z

    This document identifies 241-BY Tank Farm (BY Farm) and 241-TY Tank Farm (TY Farm) lead causes and locations for the 100 series leaking tanks (241-BY-103, 241-TY-103, 241-TY-104, 241-TY-105 and 241-TY-106) identified in RPP-RPT-43704, Hanford BY Farm Leak Assessments Report, and in RPP-RPT-42296, Hanford TY Farm Leak Assessments Report. This document satisfies the BY and TY Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  3. Hanford Single-Shell Tank Leak Causes and Locations - 241-SX Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L. [Washington River Protection Solutions (United States); Harlow, Donald G. [Washington River Protection Solutions (United States)

    2014-01-08T23:59:59.000Z

    This document identifies 241-SX Tank Farm (SX Farm) leak causes and locations for the 100 series leaking tanks (241-SX-107, 241-SX-108, 241-SX-109, 241-SX-111, 241-SX-112, 241-SX-113, 241-SX-114, and 241-SX-115) identified in RPP-ENV-39658, Rev. 0, Hanford SX-Farm Leak Assessments Report. This document satisfies the SX Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

  4. 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,

  5. 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,...

  6. 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,...

  7. 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,...

  8. 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

  9. 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...

  10. 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

  11. 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.

  12. 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...

  13. 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...

  14. 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...

  15. 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,...

  16. 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...

  17. 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...

  18. 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 ......

  19. 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...

  20. 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...

  1. 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...

  2. 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 -...

  3. 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

  4. 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...

  5. 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,...

  6. 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...

  7. 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...

  8. 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...

  9. Assessment of crack opening area for leak rates

    SciTech Connect (OSTI)

    Sharples, J.K.; Bouchard, P.J.

    1997-04-01T23:59:59.000Z

    This paper outlines the background to recommended crack opening area solutions given in a proposed revision to leak before break guidance for the R6 procedure. Comparisons with experimental and analytical results are given for some selected cases of circumferential cracks in cylinders. It is shown that elastic models can provide satisfactory estimations of crack opening displacement (and area) but they become increasingly conservative for values of L{sub r} greater than approximately 0.4. The Dugdale small scale yielding model gives conservative estimates of crack opening displacement with increasing enhancement for L{sub r} values greater than 0.4. Further validation of the elastic-plastic reference stress method for up to L{sub r} values of about 1.0 is presented by experimental and analytical comparisons. Although a more detailed method, its application gives a best estimate of crack opening displacement which may be substantially greater than small scale plasticity models. It is also shown that the local boundary conditions in pipework need to be carefully considered when evaluating crack opening area for through-wall bending stresses resulting from welding residual stresses or geometry discontinuities.

  10. Further development of an in-pipe leak detection sensor's mobility platform

    E-Print Network [OSTI]

    Moore, Frederick M

    2013-01-01T23:59:59.000Z

    Water leakage is a major global problem and smaller sized leaks are difficult to find despite their prevalence in most water distribution systems. Previous attempts to develop a mobility platform for a sensor in use in ...

  11. AIRBORNE, OPTICAL REMOTE SENSNG OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION

    SciTech Connect (OSTI)

    Jerry Myers

    2005-04-15T23:59:59.000Z

    Ophir Corporation was awarded a contract by the U. S. Department of Energy, National Energy Technology Laboratory under the Project Title ''Airborne, Optical Remote Sensing of Methane and Ethane for Natural Gas Pipeline Leak Detection'' on October 14, 2002. The scope of the work involved designing and developing an airborne, optical remote sensor capable of sensing methane and, if possible, ethane for the detection of natural gas pipeline leaks. Flight testing using a custom dual wavelength, high power fiber amplifier was initiated in February 2005. Ophir successfully demonstrated the airborne system, showing that it was capable of discerning small amounts of methane from a simulated pipeline leak. Leak rates as low as 150 standard cubic feet per hour (scf/h) were detected by the airborne sensor.

  12. BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010...

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

    40' - June 3 2010 (1 of 4) BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010 (1 of 4) Addthis Description Footage of the BP Oil Spill Duration 0:15...

  13. BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...

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

    3 of 4) BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (3 of 4) Addthis Description Footage of the BP Oil Spill Duration 0:19...

  14. BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...

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

    2 of 4) BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (2 of 4) Addthis Description Footage of the BP Oil Spill Duration 0:13...

  15. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  16. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1995-01-01T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  17. Intelligent Coatings for Location And Detection of Leaks (IntelliCLAD...

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

    danger of a gas leak. Ever since the tragic natural gas explosion of 1937 in a New London, Texas school building, various governments have mandated that odorants be added to...

  18. U.S. strategic petroleum reserve Big Hill 114 leak analysis 2012.

    SciTech Connect (OSTI)

    Lord, David L.; Roberts, Barry L.; Lord, Anna C. Snider; Sobolik, Steven Ronald; Park, Byoung Yoon; Rudeen, David Keith [GRAM, Inc., Albuquerque, NM

    2013-06-01T23:59:59.000Z

    This report addresses recent well integrity issues related to cavern 114 at the Big Hill Strategic Petroleum Reserve site. DM Petroleum Operations, M&O contractor for the U.S. Strategic Petroleum Reserve, recognized an apparent leak in Big Hill cavern well 114A in late summer, 2012, and provided written notice to the State of Texas as required by law. DM has since isolated the leak in well A with a temporary plug, and is planning on remediating both 114 A- and B-wells with liners. In this report Sandia provides an analysis of the apparent leak that includes: (i) estimated leak volume, (ii) recommendation for operating pressure to maintain in the cavern between temporary and permanent fixes for the well integrity issues, and (iii) identification of other caverns or wells at Big Hill that should be monitored closely in light of the sequence of failures there in the last several years.

  19. Risks from Past, Current, and Potential Hanford Single Shell Tank Leaks

    SciTech Connect (OSTI)

    Triplett, Mark B.; Watson, David J.; Wellman, Dawn M.

    2013-05-24T23:59:59.000Z

    Due to significant delays in constructing and operating the Waste Treatment Plant, which is needed to support retrieval of waste from Hanford’s single shell tanks (SSTs), SSTs may now be required to store tank waste for two to three more decades into the future. Many SSTs were built almost 70 years ago, and all SSTs are well beyond their design lives. Recent examination of monitoring data suggests several of the tanks, which underwent interim stabilization a decade or more ago, may be leaking small amounts (perhaps 150–300 gallons per year) to the subsurface environment. A potential leak from tank T-111 is estimated to have released approximately 2,000 gallons into the subsurface. Observations of past leak events, recently published simulation results, and new simulations all suggest that recent leaks are unlikely to affect underlying groundwater above regulatory limits. However, these recent observations remind us that much larger source terms are still contained in the tanks and are also present in the vadose zone from historical intentional and unintentional releases. Recently there have been significant improvements in methods for detecting and characterizing soil moisture and contaminant releases, understanding and controlling mass-flux, and remediating deep vadose zone and groundwater plumes. To ensure extended safe storage of tank waste in SSTs, the following actions are recommended: 1) Improve capabilities for intrusion and leak detection. 2) Develop defensible conceptual models of intrusion and leak mechanisms. 3) Apply enhanced subsurface characterization methods to improve detection and quantification of moisture changes beneath tanks. 4) Maintain a flux-based assessment of past, present, and potential tank leaks to assess risks and to maintain priorities for applying mitigation actions. 5) Implement and maintain effective mitigation and remediation actions to protect groundwater resources. These actions will enable limited resources to be applied to the most beneficial actions. A systems-based approach will support extended safe storage of tank waste, reduce the risks from tank leaks, and protect human health and the environment.

  20. Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

    SciTech Connect (OSTI)

    Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

    1985-11-01T23:59:59.000Z

    A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs.

  1. Research of documents pertaining to waste migration from leaking single-shell tanks

    SciTech Connect (OSTI)

    Consort, S.D. [ICF Kaiser Hanford Co., Richland, WA (United States)

    1994-09-30T23:59:59.000Z

    This report contains the results from an investigation of the literature concerning single-shell tank (SST) leaks on the Hanford Site. The purpose of the investigation is to determine if available data confirm or refute the assertion that leaked waste from the SSTs has reached ground water. There are 67 leaking single-shell tanks (SSTs) on the Hanford Site. Although the maximum volume of leaked waste is approximately 4,013,000 L (1,060,000 gal), it is not the only waste in the ground beneath the 200 Area. Before 1966, supernatant solution was intentionally discharged from the cascading SSTs to the ground. Other leaks from piping and surface spills contributed to the waste in the ground. The maximum estimated volume of unintentionally leaked waste from the tanks is less than 1% of the intentionally released liquid waste that was sent to the cribs and trenches from the SSTs. The volume does not include the liquid waste sent intentionally from other facilities directly to the cribs, trenches, and injection wells. The components and concentrations of the intentionally released waste were in compliance with applicable standards at the time of release.

  2. 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

  3. Leak Detection and H2 Sensor Development for Hydrogen Applications

    SciTech Connect (OSTI)

    Brosha, Eric L. [Los Alamos National Laboratory

    2012-07-10T23:59:59.000Z

    The objectives of this report are: (1) Develop a low cost, low power, durable, and reliable hydrogen safety sensor for a wide range of vehicle and infrastructure applications; (2) Continually advance test prototypes guided by materials selection, sensor design, electrochemical R&D investigation, fabrication, and rigorous life testing; (3) Disseminate packaged sensor prototypes and control systems to DOE Laboratories and commercial parties interested in testing and fielding advanced prototypes for cross-validation; (4) Evaluate manufacturing approaches for commercialization; and (5) Engage an industrial partner and execute technology transfer. Recent developments in the search for sustainable and renewable energy coupled with the advancements in fuel cell powered vehicles (FCVs) have augmented the demand for hydrogen safety sensors. There are several sensor technologies that have been developed to detect hydrogen, including deployed systems to detect leaks in manned space systems and hydrogen safety sensors for laboratory and industrial usage. Among the several sensing methods electrochemical devices that utilize high temperature-based ceramic electrolytes are largely unaffected by changes in humidity and are more resilient to electrode or electrolyte poisoning. The desired sensing technique should meet a detection threshold of 1% (10,000 ppm) H{sub 2} and response time of {approx_equal}1 min, which is a target for infrastructure and vehicular uses. Further, a review of electrochemical hydrogen sensors by Korotcenkov et.al and the report by Glass et.al suggest the need for inexpensive, low power, and compact sensors with long-term stability, minimal cross-sensitivity, and fast response. This view has been largely validated and supported by the fuel cell and hydrogen infrastructure industries by the NREL/DOE Hydrogen Sensor Workshop held on June 8, 2011. Many of the issues preventing widespread adoption of best-available hydrogen sensing technologies available today outside of cost, derive from excessive false positives and false negatives arising from signal drift and unstable sensor baseline; both of these problems necessitate the need for unacceptable frequent calibration.

  4. A LOW-COST GPR GAS PIPE & LEAK DETECTOR

    SciTech Connect (OSTI)

    David Cist; Alan Schutz

    2005-03-30T23:59:59.000Z

    A light-weight, easy to use ground penetrating radar (GPR) system for tracking metal/non-metal pipes has been developed. A pre-production prototype instrument has been developed whose production cost and ease of use should fit important market niches. It is a portable tool which is swept back and forth like a metal detector and which indicates when it goes over a target (metal, plastic, concrete, etc.) and how deep it is. The innovation of real time target detection frees the user from having to interpret geophysical data and instead presents targets as dots on the screen. Target depth is also interpreted automatically, relieving the user of having to do migration analysis. In this way the user can simply walk around looking for targets and, by ''connecting the dots'' on the GPS screen, locate and follow pipes in real time. This is the first tool known to locate metal and non-metal pipes in real time and map their location. This prototype design is similar to a metal detector one might use at the beach since it involves sliding a lightweight antenna back and forth over the ground surface. The antenna is affixed to the end of an extension that is either clipped to or held by the user. This allows him to walk around in any direction, either looking for or following pipes with the antenna location being constantly recorded by the positioning system. Once a target appears on the screen, the user can locate by swinging the unit to align the cursor over the dot. Leak detection was also a central part of this project, and although much effort was invested into its development, conclusive results are not available at the time of the writing of this document. Details of the efforts that were made as a part of this cooperative agreement are presented.

  5. RADIATION DAMAGE OF GERMANIUM DETECTORS

    E-Print Network [OSTI]

    Pehl, Richard H.

    2011-01-01T23:59:59.000Z

    the high-energy proton damage than was the planar detector.as far as radiation damage is concerned. Unfortunately, some28-29, 1978 LBL-7967 RADIATION DAMAGE OF GERMANIUM DETECTORS

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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. ...

  15. 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.

  16. Composite heat damage assessment

    SciTech Connect (OSTI)

    Janke, C.J.; Wachter, E.A. [Oak Ridge National Lab., TN (United States); Philpot, H.E. [Oak Ridge K-25 Site, TN (United States); Powell, G.L. [Oak Ridge Y-12 Plant, TN (United States)

    1993-12-31T23:59:59.000Z

    The effects of heat damage were determined on the residual mechanical, physical, and chemical properties of IM6/3501-6 laminates, and potential nondestructive techniques to detect and assess material heat damage were evaluated. About one thousand preconditioned specimens were exposed to elevated temperatures, then cooled to room temperature and tested in compression, flexure, interlaminar shear, shore-D hardness, weight loss, and change in thickness. Specimens experienced significant and irreversible reduction in their residual properties when exposed to temperatures exceeding the material upper service temperature of this material (350{degrees}F). The Diffuse Reflectance Infrared Fourier Transform and Laser-Pumped Fluorescence techniques were found to be capable of rapid, in-service, nondestructive detection and quantitation of heat damage in IM6/3501- 6. These techniques also have the potential applicability to detect and assess heat damage effects in other polymer matrix composites.

  17. Controlling Beaver Damage

    E-Print Network [OSTI]

    Texas Wildlife Services

    2007-03-13T23:59:59.000Z

    Beavers are important because their dams stabilize creek flow, slow runoff and create ponds. However, these same dams can negatively alter the flow of creeks. Damage prevention, control and various trapping methods are discussed in this publication....

  18. Controlling Opossum Damage

    E-Print Network [OSTI]

    Texas Wildlife Services

    2007-05-23T23:59:59.000Z

    damage; however, their pelts can be sold only during the furbearer season and with the proper licenses. Other furbearers include beaver, otter, mink, nutria, ringtailed cat, badger, skunk, weasel, raccoon, muskrat, fox and civet cat. Homeowners...

  19. 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

  20. 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.

  1. A UNIFIED FAILURE/DAMAGE APPROACH TO BATTLE DAMAGE REGENERATION

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    A UNIFIED FAILURE/DAMAGE APPROACH TO BATTLE DAMAGE REGENERATION : APPLICATION TO GROUND MILITARY-availability. Military weapon systems availability can be affected by system failures or by damage to the system damage into account in their more general dependability studies. This paper takes a look at the issues

  2. Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture

    SciTech Connect (OSTI)

    Walston, S; Rowland, M; Campbell, K

    2011-07-27T23:59:59.000Z

    It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting in a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.

  3. Acceptance test report for the AN valve pit leak detection and low point drain assembly mock up test procedure

    SciTech Connect (OSTI)

    EWER, K.L.

    1999-07-20T23:59:59.000Z

    This document describes The Performance Mock-up Test Procedure for the Valve Pit Leak Detection and Low Point Drain Assembly Performance Mock-Up Test Procedure.

  4. Influence of wetting effect at the outer surface of the pipe on increase in leak rate - experimental results and discussion

    SciTech Connect (OSTI)

    Isozaki, Toshikuni; Shibata, Katsuyuki

    1997-04-01T23:59:59.000Z

    Experimental and computed results applicable to Leak Before Break analysis are presented. The specific area of investigation is the effect of the temperature distribution changes due to wetting of the test pipe near the crack on the increase in the crack opening area and leak rate. Two 12-inch straight pipes subjected to both internal pressure and thermal load, but not to bending load, are modelled. The leak rate was found to be very susceptible to the metal temperature of the piping. In leak rate tests, therefore, it is recommended that temperature distribution be measured precisely for a wide area.

  5. Leak before break evaluation for main steam piping system made of SA106 Gr.C

    SciTech Connect (OSTI)

    Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

    1997-04-01T23:59:59.000Z

    The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

  6. Flight Testing of an Advanced Airborne Natural Gas Leak Detection System

    SciTech Connect (OSTI)

    Dawn Lenz; Raymond T. Lines; Darryl Murdock; Jeffrey Owen; Steven Stearns; Michael Stoogenke

    2005-10-01T23:59:59.000Z

    ITT Industries Space Systems Division (Space Systems) has developed an airborne natural gas leak detection system designed to detect, image, quantify, and precisely locate leaks from natural gas transmission pipelines. This system is called the Airborne Natural Gas Emission Lidar (ANGEL) system. The ANGEL system uses a highly sensitive differential absorption Lidar technology to remotely detect pipeline leaks. The ANGEL System is operated from a fixed wing aircraft and includes automatic scanning, pointing system, and pilot guidance systems. During a pipeline inspection, the ANGEL system aircraft flies at an elevation of 1000 feet above the ground at speeds of between 100 and 150 mph. Under this contract with DOE/NETL, Space Systems was funded to integrate the ANGEL sensor into a test aircraft and conduct a series of flight tests over a variety of test targets including simulated natural gas pipeline leaks. Following early tests in upstate New York in the summer of 2004, the ANGEL system was deployed to Casper, Wyoming to participate in a set of DOE-sponsored field tests at the Rocky Mountain Oilfield Testing Center (RMOTC). At RMOTC the Space Systems team completed integration of the system and flew an operational system for the first time. The ANGEL system flew 2 missions/day for the duration for the 5-day test. Over the course of the week the ANGEL System detected leaks ranging from 100 to 5,000 scfh.

  7. 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.

  8. EXTENDED PERFORMANCE HANDHELD AND MOBILE SENSORS FOR REMOTE DETECTION OF NATURAL GAS LEAKS

    SciTech Connect (OSTI)

    Michael B. Frish; B. David Green; Richard T. Wainner; Francesca Scire-Scappuzzo; Paul Cataldi; Matthew C. Laderer

    2005-05-01T23:59:59.000Z

    This report summarizes work performed by Physical Sciences Inc. (PSI) to advance the state-of-the-art of surveying for leaks of natural gas from transmission and distribution pipelines. The principal project goal was to develop means of deploying on an automotive platform an improved version of the handheld laser-based standoff natural gas leak detector previously developed by PSI and known as the Remote Methane Leak Detector or RMLD. A laser beam which interrogates the air for methane is projected from a spinning turret mounted upon a van. As the van travels forward, the laser beam scans an arc to the front and sides of the van so as to survey across streets and to building walls from a moving vehicle. When excess methane is detected within the arc, an alarm is activated. In this project, we built and tested a prototype Mobile RMLD (MRMLD) intended to provide lateral coverage of 10 m and one lateral scan for every meter of forward motion at forward speeds up to 10 m/s. Using advanced detection algorithms developed as part of this project, the early prototype MRMLD, installed on the back of a truck, readily detected simulated gas leaks of 50 liters per hour. As a supplement to the originally planned project, PSI also participated in a DoE demonstration of several gas leak detection systems at the Rocky Mountain Oilfield Testing Center (RMOTC) during September 2004. Using a handheld RMLD upgraded with the advanced detection algorithms developed in this project, from within a moving vehicle we readily detected leaks created along the 7.4 mile route of a virtual gas transmission pipeline.

  9. Estimation of Leak Rate from the Emergency Pump Well in L-Area Complex Basin

    SciTech Connect (OSTI)

    Duncan, A

    2005-12-19T23:59:59.000Z

    This report provides an estimate of the leak rate from the emergency pump well in L-basin that is to be expected during an off-normal event. This estimate is based on expected shrinkage of the engineered grout (i.e., controlled low strength material) used to fill the emergency pump well and the header pipes that provide the dominant leak path from the basin to the lower levels of the L-Area Complex. The estimate will be used to provide input into the operating safety basis to ensure that the water level in the basin will remain above a certain minimum level. The minimum basin water level is specified to ensure adequate shielding for personnel and maintain the ''as low as reasonably achievable'' concept of radiological exposure. The need for the leak rate estimation is the existence of a gap between the fill material and the header pipes, which penetrate the basin wall and would be the primary leak path in the event of a breach in those pipes. The gap between the pipe and fill material was estimated based on a full scale demonstration pour that was performed and examined. Leak tests were performed on full scale pipes as a part of this examination. Leak rates were measured to be on the order of 0.01 gallons/minute for completely filled pipe (vertically positioned) and 0.25 gallons/minute for partially filled pipe (horizontally positioned). This measurement was for water at 16 feet head pressure and with minimal corrosion or biofilm present. The effect of the grout fill on the inside surface biofilm of the pipes is the subject of a previous memorandum.

  10. An upgraded heat transfer fluid eliminates odors and leaks

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    At Morton, persistent leakage of an aromatics-based heat transfer fluid left its mark--a black, oxidized residue at flange and valve locations. By switching to a high-purity fluid from a paraffinic hydrocarbon base stock, the firm eliminated odors and sticky residue, and improved heat transfer. After four years of operation with the paraffinic heat transfer fluid, Morton continues to have no odor problems and virtually no flange or packing leakage. As an added bonus, the heat transfer coefficient of the new fluid allows Morton to operate the systems 10--15 F cooler than when the company used the traditional, aromatic fluid. This has cut fuel use and reduced the potential for thermal damage to the heat transfer fluid, process fluid and process equipment.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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,

  16. 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:

  17. Development of a cold cathode ion source for a mass spectrometer type vacuum leak detector

    E-Print Network [OSTI]

    Thomas, Harold Albert

    1947-01-01T23:59:59.000Z

    of about 7 cm., discharge voltage of about 2000 volts, discharge current of 10 ma., and a magnetic field strength of approximately 2200 Oersteds. As a leak detector it had a differential sensitivity of one part of helium in 10,000 parts of air ? about... for this ion souree as for the first type tested# Be? cause of the simpler construction and fewer components reauired, it appears that this type of source would have some valuable possibili? ties as a mass spectrometer ion source for the leak detector...

  18. 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.

  19. 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.

  20. 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.

  1. 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).

  2. SEVENTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE

    SciTech Connect (OSTI)

    Daugherty, W.

    2012-08-30T23:59:59.000Z

    A series of experiments to monitor the aging performance of Viton® GLT O-rings used in the Model 9975 package has been ongoing since 2004 at the Savannah River National Laboratory. Seventy tests using mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 ºF. They were leak-tested initially and have been tested periodically to determine if they meet the criterion of leak-tightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 ºF. High temperature aging continues for 23 GLT O-ring fixtures at 200 – 270 ºF. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 ºF and higher temperatures, and in 8 fixtures aging at 300 ºF. The remaining GLT O-ring fixtures aging at 300 ºF have been retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 ºF for 54-72 months, which is still bounding to O-ring temperatures during storage in K-Area Complex (KAC). Based on expectations that the fixtures aging at 200 ºF will remain leak-tight for a significant period yet to come, 2 additional fixtures began aging in 2011 at an intermediate temperature of 270 ºF, with hopes that they may reach a failure condition before the 200 ºF fixtures. High temperature aging continues for 6 GLT-S O-ring fixtures at 200 – 300 ºF. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 ºF. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 - 300 ºF for 30 - 36 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the O-rings displayed a compression set ranging from 51 – 96%. This is greater than seen to date for any packages inspected during KAC field surveillance (24% average). For GLT O-rings, separate service life estimates have been made based on the O-ring fixture leak test data and based on compression stress relaxation (CSR) data. These two predictive models show reasonable agreement at higher temperatures (350 – 400 ºF). However, at 300 ºF, the room temperature leak test failures to date experienced longer aging times than predicted by the CSRbased model. This suggests that extrapolations of the CSR model predictions to temperatures below 300 ºF will provide a conservative prediction of service life relative to the leak rate criterion. Leak test failure data at lower temperatures are needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining PCV O-ring fixtures.

  3. Heat Leak into Cryostat #1 through 304SS or G10 Supports Robert J. Weggel, Magnet Optimization Research Engineering, LLC

    E-Print Network [OSTI]

    McDonald, Kirk

    Heat Leak into Cryostat #1 through 304SS or G10 Supports Robert J. Weggel, Magnet Optimization for refrigeration to cope with the heat leak through mechanical supports of Type 304 stainless steel (SS) with warm times the cross section of the SS support) requires only 18% as much power for refrigeration

  4. Efficient model-based leak detection in boiler steam-water systems Xi Sun, Tongwen Chen *, Horacio J. Marquez

    E-Print Network [OSTI]

    Marquez, Horacio J.

    Efficient model-based leak detection in boiler steam-water systems Xi Sun, Tongwen Chen *, Horacio detection in boiler steam-water systems. The algorithm has been tested using real industrial data from Syncrude Canada, and has proven to be effective in detection of boiler tube or steam leaks; proper

  5. Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

    SciTech Connect (OSTI)

    Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

    1986-10-01T23:59:59.000Z

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.

  6. 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.

  7. 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.

  8. 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...

  9. 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,...

  10. 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...

  11. 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...

  12. 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...

  13. 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.

  14. 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...

  15. 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...

  16. 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...

  17. Unaccounted-for gas project. Leak Task Force. Volume 4. Final report

    SciTech Connect (OSTI)

    Cowgill, R.M.; Robertson, J.L.; Grinstead, J.R.; Luttrell, D.J.; Walden, E.R.

    1990-06-07T23:59:59.000Z

    The study was aimed at determining unaccounted-for (UAF) gas volumes resulting from operating Pacific Gas and Electric Co.'s transmission and distribution systems during 1987. The Leak Task Force quantified unintentional gas losses (leakage and dig-ins). Results show that 1987 gas leakage accounted for less than 5% of the operating UAF.

  18. Problem Type Problem Type Description Air Conditioning Air conditioner not working, leaking, etc

    E-Print Network [OSTI]

    Tennessee, University of

    Problem Type Problem Type Description Air Conditioning Air conditioner not working, leaking, etc, Microfridges Doors and Hardware Door repair/replace Lock, latch or hinge repair, key stuck; Lost or stolen key, repair or replace Shades/Blinds Window treatment - repair or replace Washer/Dryer Washer/Dryer repair

  19. UTILITIES PROBLEMS AND FAILURES Electrical or plumbing failure/Flooding/Water leak/Natural gas

    E-Print Network [OSTI]

    Fernandez, Eduardo

    UTILITIES PROBLEMS AND FAILURES Electrical or plumbing failure/Flooding/Water leak/Natural gas for electrical shock. NOTIFY University Police. What should I do if I smell natural or propane gas? LEAVE/Repair line, 7-6333, or CALL the Campus University Police or Security at (561) 297-3500 or 911

  20. UTILITIES PROBLEMS AND FAILURES ELECTRICAL OR PLUMBING FAILURE/FLOODING/WATER LEAK

    E-Print Network [OSTI]

    Fernandez, Eduardo

    UTILITIES PROBLEMS AND FAILURES ELECTRICAL OR PLUMBING FAILURE/FLOODING/WATER LEAK NATURAL GAS - F 8a - 5p HBOI@FAU Security (772) 216-1124 Afterhours, Weekends or Holidays What should I do Police 911. · NOTIFY Building Safety personnel when possible. What should I do if I smell natural

  1. UTILITIES PROBLEMS AND FAILURES Electrical or plumbing failure/Flooding/Water leak/Natural gas or

    E-Print Network [OSTI]

    Fernandez, Eduardo

    UTILITIES PROBLEMS AND FAILURES Electrical or plumbing failure/Flooding/Water leak/Natural gas Physical Plant (772) 242-2246 M - F 8a - 5p (954) 762-5040 HBOI@FAU Security (772) 216-1124 Afterhours University Police. NOTIFY Building Safety personnel when possible. What should I do if I smell natural

  2. Detecting leak regions through model falsification GAUDENZ MOSER AND IAN F.C. SMITH

    E-Print Network [OSTI]

    Candea, George

    management support. Since a significant percentage of fresh water is lost globally due to leaks in these networks, the challenge to improve performance is compatible with goals of sustainable development is a precious resource that is necessary to preserve. Preservation involves reducing losses in the water

  3. Strontium and cesium radionuclide leak detection alternatives in a capsule storage pool

    SciTech Connect (OSTI)

    Larson, D.E.; Crawford, T.W.; Joyce, S.M.

    1981-08-01T23:59:59.000Z

    A study was performed to assess radionuclide leak-detection systems for use in locating a capsule leaking strontium-90 or cesium-137 into a water-filled pool. Each storage pool contains about 35,000 L of water and up to 715 capsules, each of which contains up to 150 kCi strontium-90 or 80 kCi cesium-137. Potential systems assessed included instrumental chemical analyses, radionuclide detection, visual examination, and other nondestructive nuclear-fuel examination techniques. Factors considered in the assessment include: cost, simplicity of maintenance and operation, technology availability, reliability, remote operation, sensitivity, and ability to locate an individual leaking capsule in its storage location. The study concluded that an adaption of the spent nuclear-fuel examination technique of wet sipping be considered for adaption. In the suggested approoch, samples would be taken continuously from pool water adjacent to the capsule(s) being examined for remote radiation detection. In-place capsule isolation and subsequent water sampling would confirm that a capsule was leaking radionuclides. Additional studies are needed before implementing this option. Two other techniques that show promise are ultrasonic testing and eddy-current testing.

  4. Tank 241-AY-102 Leak Assessment Supporting Documentation: Miscellaneous Reports, Letters, Memoranda, And Data

    SciTech Connect (OSTI)

    Engeman, J. K.; Girardot, C. L.; Harlow, D. G.; Rosenkrance, C. L.

    2012-12-20T23:59:59.000Z

    This report contains reference materials cited in RPP-ASMT -53793, Tank 241-AY-102 Leak Assessment Report, that were obtained from the National Archives Federal Records Repository in Seattle, Washington, or from other sources including the Hanford Site's Integrated Data Management System database (IDMS).

  5. Oil spill nears the beaches of Florida, and the leak may not be plugged before Christmas

    E-Print Network [OSTI]

    Belogay, Eugene A.

    Oil spill nears the beaches of Florida, and the leak may not be plugged before Christmas By David Gardner Last updated at 11:32 AM on 3rd June 2010 BP's giant oil slick was bearing down on Florida holidaymakers a year visit Florida and state leaders fear the oil will devastate a tourist industry

  6. Rigorous Simulation of Accidental Leaks from High-Pressure Storage Vessels

    E-Print Network [OSTI]

    Alisha, -

    2014-07-07T23:59:59.000Z

    of nature. The released chemical can form and disperse as vapor cloud leading to fire, explosion, or toxic exposure. The resulting leak could be single phase or multiphase release, choked or non-choked. These releases could result in liquid spills, vapor...

  7. Rigorous Simulation of Accidental Leaks from High-Pressure Storage Vessels 

    E-Print Network [OSTI]

    Alisha, -

    2014-07-07T23:59:59.000Z

    of nature. The released chemical can form and disperse as vapor cloud leading to fire, explosion, or toxic exposure. The resulting leak could be single phase or multiphase release, choked or non-choked. These releases could result in liquid spills, vapor...

  8. 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

  9. 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.

  10. FIFTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE

    SciTech Connect (OSTI)

    Daugherty, W.; Hoffman, E.

    2010-11-01T23:59:59.000Z

    A series of experiments to monitor the aging performance of Viton{sup reg.} GLT O-rings used in the Model 9975 package has been ongoing for six years at the Savannah River National Laboratory. Sixty-seven mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested at nominal six month intervals to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 36 GLT O-ring fixtures at 200--350 F. Room temperature leak test failures have been experienced in 5 of the GLT O-ring fixtures aging at 300 and 350 F, and in all 3 of the GLT O-ring fixtures aging at higher temperatures. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 30--48 months, which is still bounding to O-ring temperatures during storage in KAMS. High temperature aging continues for 6 GLT-S O-ring fixtures at 200--300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 or 300 F for 19 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the O-rings displayed a compression set ranging from 51--95%. This is significantly greater than seen to date for packages inspected during KAMS field surveillance (23% average). For GLT O-rings, service life based on the room temperature leak rate criterion is comparable to that predicted by compression stress relaxation (CSR) data at higher temperatures (350--400 F). While there are no comparable failure data yet at aging temperatures below 300 F, extrapolations of the data for GLT O-rings suggests that CSR model predictions provide a conservative prediction of service life relative to the leak rate criterion. Failure data at lower temperatures is needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining fixtures.

  11. 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

  12. 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...

  13. 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

  14. 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.

  15. Engineering evaluation of alternatives: Managing the assumed leak from single-shell Tank 241-T-101

    SciTech Connect (OSTI)

    Brevick, C.H. [ICF Kaiser Hanford Co., Richland, WA (United States); Jenkins, C. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-02-01T23:59:59.000Z

    At mid-year 1992, the liquid level gage for Tank 241-T-101 indicated that 6,000 to 9,000 gal had leaked. Because of the liquid level anomaly, Tank 241-T-101 was declared an assumed leaker on October 4, 1992. SSTs liquid level gages have been historically unreliable. False readings can occur because of instrument failures, floating salt cake, and salt encrustation. Gages frequently self-correct and tanks show no indication of leak. Tank levels cannot be visually inspected and verified because of high radiation fields. The gage in Tank 241-T-101 has largely corrected itself since the mid-year 1992 reading. Therefore, doubt exists that a leak has occurred, or that the magnitude of the leak poses any immediate environmental threat. While reluctance exists to use valuable DST space unnecessarily, there is a large safety and economic incentive to prevent or mitigate release of tank liquid waste into the surrounding environment. During the assessment of the significance of the Tank 241-T-101 liquid level gage readings, Washington State Department of Ecology determined that Westinghouse Hanford Company was not in compliance with regulatory requirements, and directed transfer of the Tank 241-T-101 liquid contents into a DST. Meanwhile, DOE directed WHC to examine reasonable alternatives/options for safe interim management of Tank 241-T-101 wastes before taking action. The five alternatives that could be used to manage waste from a leaking SST are: (1) No-Action, (2) In-Tank Stabilization, (3) External Tank Stabilization, (4) Liquid Retrieval, and (5) Total Retrieval. The findings of these examinations are reported in this study.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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 ],

  8. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1

    SciTech Connect (OSTI)

    Jo, J.; Lin, C.C.; Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1995-05-01T23:59:59.000Z

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6.

  9. Radiation damage evolution in ceramics. | EMSL

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

    Radiation damage evolution in ceramics. Radiation damage evolution in ceramics. Abstract: A review is presented of recent results on radiation damage production, defect...

  10. Regulation with anticipated learning about environmental damages

    E-Print Network [OSTI]

    Karp, L; Zhang, J

    2006-01-01T23:59:59.000Z

    abatement costs and environmental damages, and a generalemissions. 2.2 Environmental damages and learning Let S t begas stocks and environmental damages. In some respects these

  11. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01T23:59:59.000Z

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  12. Real time imaging of live cell ATP leaking or release events by chemiluminescence microscopy

    SciTech Connect (OSTI)

    Zhang, Yun

    2008-12-18T23:59:59.000Z

    The purpose of this research was to expand the chemiluminescence microscopy applications in live bacterial/mammalian cell imaging and to improve the detection sensitivity for ATP leaking or release events. We first demonstrated that chemiluminescence (CL) imaging can be used to interrogate single bacterial cells. While using a luminometer allows detecting ATP from cell lysate extracted from at least 10 bacterial cells, all previous cell CL detection never reached this sensitivity of single bacteria level. We approached this goal with a different strategy from before: instead of breaking bacterial cell membrane and trying to capture the transiently diluted ATP with the firefly luciferase CL assay, we introduced the firefly luciferase enzyme into bacteria using the modern genetic techniques and placed the CL reaction substrate D-luciferin outside the cells. By damaging the cell membrane with various antibacterial drugs including antibiotics such as Penicillins and bacteriophages, the D-luciferin molecules diffused inside the cell and initiated the reaction that produces CL light. As firefly luciferases are large protein molecules which are retained within the cells before the total rupture and intracellular ATP concentration is high at the millmolar level, the CL reaction of firefly luciferase, ATP and D-luciferin can be kept for a relatively long time within the cells acting as a reaction container to generate enough photons for detection by the extremely sensitive intensified charge coupled device (ICCD) camera. The result was inspiring as various single bacterium lysis and leakage events were monitored with 10-s temporal resolution movies. We also found a new way of enhancing diffusion D-luciferin into cells by dehydrating the bacteria. Then we started with this novel single bacterial CL imaging technique, and applied it for quantifying gene expression levels from individual bacterial cells. Previous published result in single cell gene expression quantification mainly used a fluorescence method; CL detection is limited because of the difficulty to introduce enough D-luciferin molecules. Since dehydration could easily cause proper size holes in bacterial cell membranes and facilitate D-luciferin diffusion, we used this method and recorded CL from individual cells each hour after induction. The CL light intensity from each individual cell was integrated and gene expression levels of two strain types were compared. Based on our calculation, the overall sensitivity of our system is already approaching the single enzyme level. The median enzyme number inside a single bacterium from the higher expression strain after 2 hours induction was quantified to be about 550 molecules. Finally we imaged ATP release from astrocyte cells. Upon mechanical stimulation, astrocyte cells respond by increasing intracellular Ca{sup 2+} level and releasing ATP to extracellular spaces as signaling molecules. The ATP release imaged by direct CL imaging using free firefly luciferase and D-luciferin outside cells reflects the transient release as well as rapid ATP diffusion. Therefore ATP release detection at the cell surface is critical to study the ATP release mechanism and signaling propagation pathway. We realized this cell surface localized ATP release imaging detection by immobilizing firefly luciferase to streptavidin beads that attached to the cell surface via streptavidin-biotin interactions. Both intracellular Ca{sup 2+} propagation wave and extracellular ATP propagation wave at the cell surface were recorded with fluorescence and CL respectively. The results imply that at close distances from the stimulation center (<120 {micro}m) extracellular ATP pathway is faster, while at long distances (>120 {micro}m) intracellular Ca{sup 2+} signaling through gap junctions seems more effective.

  13. 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.

  14. 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.

  15. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    SciTech Connect (OSTI)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01T23:59:59.000Z

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  16. 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

  17. 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

  18. 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

  19. Assessment of possible consequences of a hypothetical reactivity accident associated with a {open_quotes}Topaz-2{close_quotes} spacecraft reactor entering water

    SciTech Connect (OSTI)

    Glushkov, E.S.; Ermoshin, M.Yu.; Ponomarev-Stepnoi; Skorlygin, V.V.

    1994-12-01T23:59:59.000Z

    An accident analysis for a Russian Topaz-2 nuclear reactor is summarized. The accident scenario involves emergency return from orbit, severe damage to reactor structural elements, and subsequent falling of the reactor core into the ocean. The thermionic converter reactor, used in spacecraft, has a large neutron leakage which decreases when water enters the inner core cavity. Preliminary results of numerical modeling, summarized in the article, show that the possible consequences of the hypothetical accidental submersion are limited. 8 refs., 2 figs., 2 tabs.

  20. 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.

  1. 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.

  2. 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.

  3. Is the situation and immediate threat to life and health? Spill/Leak/Release Medical Emergency Fire or Flammable Gas Spill/Leak/Release Medical Emergency Fire or Flammable Gas Chemical Odor? Possible Fire / Natural Gas

    E-Print Network [OSTI]

    ? Possible Fire / Natural Gas (including chemicals and bio agents") (not including chemicals or bio agents Fire or Flammable Gas Spill/Leak/Release Medical Emergency Fire or Flammable Gas Chemical Odor

  4. 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.

  5. Hanford Double-Shell Tank AY-102 Radioactive Waste Leak Investigation Update - 15302

    SciTech Connect (OSTI)

    Washenfelder, D. J.; Johnson, J. M.

    2014-12-22T23:59:59.000Z

    Tank AY-102 was the first of 28 double-shell radioactive waste storage tanks constructed at the U. S. Department of Energy’s Hanford Site, near Richland, WA. The tank was completed in 1970, and entered service in 1971. In August, 2012, an accumulation of material was discovered at two sites on the floor of the annulus that separates the primary tank from the secondary liner. The material was sampled and determined to originate from the primary tank. This paper summarizes the changes in leak behavior that have occurred during the past two years, inspections to determine the capability of the secondary liner to continue safely containing the leakage, and the initial results of testing to determine the leak mechanism.

  6. The probability of intersystem LOCA: impact due to leak testing and operational changes. Technical report

    SciTech Connect (OSTI)

    Rubin, M.P.

    1980-05-01T23:59:59.000Z

    The Reactor Safety Study (WASH-1400) identified the potential intersystem LOCA in a pressurized water reactor as a significant contributor to the risk resulting from core melt. Similar scenarios are also possible in boiling water reactors. This report evaluates various pressure isolation valve configurations used in reactors to determine the probability of intersystem LOCA. It is shown that periodic leak testing of these valves can substantially reduce intersystem LOCA probability. Specific analyses of the high pressure/low pressure interfaces in the Sequoyah (PWR) and Alan B. Barton (BWR) plants show that periodic leak testing of the pressure isolation check valves will reduce the intersystem LOCA probability to below 0.000001 per year.

  7. Advanced conceptual design report: T Plant secondary containment and leak detection upgrades. Project W-259

    SciTech Connect (OSTI)

    Hookfin, J.D.

    1995-05-12T23:59:59.000Z

    The T Plant facilities in the 200-West Area of the Hanford site were constructed in the early 1940s to produce nuclear materials in support of national defense activities. T Plant includes the 271-T facility, the 221-T facility, and several support facilities (eg, 2706-T), utilities, and tanks/piping systems. T Plant has been recommended as the primary interim decontamination facility for the Hanford site. Project W-259 will provide capital upgrades to the T Plant facilities to comply with Federal and State of Washington environmental regulations for secondary containment and leak detection. This document provides an advanced conceptual design concept that complies with functional requirements for the T Plant Secondary Containment and Leak Detection upgrades.

  8. 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)

  9. ENVIRONMENTAL MONITORING OF LEAKS USING TIME LAPSED LONG ELECTRODE ELECTRICAL RESISTIVITY

    SciTech Connect (OSTI)

    MYERS DA; RUCKER DF; FINK JB; LOKE MH

    2009-12-16T23:59:59.000Z

    Highly industrialized areas pose challenges for surface electrical resistivity characterization due to metallic infrastructure. The infrastructure is typically more conductive than the desired targets and will mask the deeper subsurface information. These challenges may be minimized if steel-cased wells are used as long electrodes in the area near the target. We demonstrate a method of using long electrodes to electrically monitor a simulated leak from an underground storage tank with both synthetic examples and a field demonstration. The synthetic examples place a simple target of varying electrical properties beneath a very low resistivity layer. The layer is meant to replicate the effects of infrastructure. Both surface and long electrodes are tested on the synthetic domain. The leak demonstration for the field experiment is simulated by injecting a high conductivity fluid in a perforated well within the S tank farm at Hanford, and the resistivity measurements are made before and after the leak test. All data are processed in four dimensions, where a regularization procedure is applied in both the time and space domains. The synthetic test case shows that the long electrode ERM could detect relative changes in resistivity that are commensurate with the differing target properties. The surface electrodes, on the other hand, had a more difficult time matching the original target's footprint. The field results shows a lowered resistivity feature develop south of the injection site after cessation of the injections. The time lapsed regularization parameter has a strong influence on the differences in inverted resistivity between the pre and post injection datasets, but the interpretation of the target is consistent across all values of the parameter. The long electrode ERM method may provide a tool for near real-time monitoring of leaking underground storage tanks.

  10. Use of the Niyama Criterion To Predict Shrinkage-Related Leaks in High-Nickel

    E-Print Network [OSTI]

    Beckermann, Christoph

    1 Use of the Niyama Criterion To Predict Shrinkage-Related Leaks in High-Nickel Steel and Nickel by the present authors determined that Nymacro = 1.0 (°C-s)1/2 /mm for nickel-based alloys M30C, M35-1 and CW12MW-shrinkage in high-nickel alloys by determining Nymicro. This is accomplished by performing metallographic analyses

  11. Location of Leaks in Pressure Testable Direct Burial Steam Distribution Conduits

    E-Print Network [OSTI]

    Sittel, M. G.; Messock, R. K.

    are ~xcavated for repair. We have successfully used this system at several locations, and in a variety of soil conditions. Tracer gas leak testing provides an effective and inexpensive method to evaluate underground conduit systems. Performed on a regular... containing all equipment to accomplish the testing is also described. ~iAACERGAS Sulfur hexaflouride (SF6) has been chosen as a tracer gas because: a) it is chemically inert, non-toxic, and has negligible pollution potential; b) it is highly...

  12. Leaking Interleavers for UEP Turbo Codes Abdul Wakeel, David Kronmueller, Werner Henkel, and Humberto Beltr~ao Neto

    E-Print Network [OSTI]

    Henkel, Werner

    Leaking Interleavers for UEP Turbo Codes Abdul Wakeel, David Kronmueller, Werner Henkel to Turbo coding's exceptional performance. An interleaver provides bit-permutation designed to ensure deterministic randomness. When applying interleavers to unequal error protecting (UEP) Turbo codes, typically

  13. Discovery of the First Leaking Double-Shell Tank - Hanford Tank 241-AY-102

    SciTech Connect (OSTI)

    Harrington, Stephanie J. [Washington River Protection Systems, Richland, WA (United States); Sams, Terry L. [Washington River Protection Systems, Richland, WA (United States)

    2013-11-06T23:59:59.000Z

    A routine video inspection of the annulus space between the primary tank and secondary liner of double-shell tank 241-AY-102 was performed in August 2012. During the inspection, unexpected material was discovered. A subsequent video inspection revealed additional unexpected material on the opposite side of the tank, none of which had been observed during inspections performed in December 2006 and January 2007. A formal leak assessment team was established to review the tank's construction and operating histories, and preparations for sampling and analysis began to determine the material's origin. A new sampling device was required to collect material from locations that were inaccessible to the available sampler. Following its design and fabrication, a mock-up test was performed for the new sampling tool to ensure its functionality and capability of performing the required tasks. Within three months of the discovery of the unexpected material, sampling tools were deployed, material was collected, and analyses were performed. Results indicated that some of the unknown material was indicative of soil, whereas the remainder was consistent with tank waste. This, along with the analyses performed by the leak assessment team on the tank's construction history, lead to the conclusion that the primary tank was leaking into the annulus. Several issues were encountered during the deployment of the samplers into the annulus. As this was the first time samples had been required from the annulus of a double-shell tank, a formal lessons learned was created concerning designing equipment for unique purposes under time constraints.

  14. Multi Canister Overpack (MCO) Combustible Gas Management Leak Test Acceptance Criteria (OCRWM)

    SciTech Connect (OSTI)

    SHERRELL, D.L.

    2000-10-10T23:59:59.000Z

    The purpose of this document is to support the Spent Nuclear Fuel Project's combustible gas management strategy while avoiding the need to impose any requirements for oxygen free atmospheres within storage tubes that contain multi-canister overpacks (MCO). In order to avoid inerting requirements it is necessary to establish and confirm leak test acceptance criteria for mechanically sealed and weld sealed MCOs that are adequte to ensure that, in the unlikely event the leak test results for any MCO were to approach either of those criteria, it could still be handled and stored in stagnant air without compromising the SNF Project's overall strategy to prevent accumulation of combustible gas mixtures within MCOs or within their surroundings. To support that strategy, this document: (1) establishes combustible gas management functions and minimum functional requirements for the MCO's mechanical seals and closure weld(s); (2) establishes a maximum practical value for the minimum required initial MCO inert backfill gas pressure; and (3) based on items 1 and 2, establishes and confirms leak test acceptance criteria for the MCO's mechanical seal and final closure weld(s).

  15. ENVIRONMENTAL MONITORING OF LEAKS USING TIME LAPSED LONG ELECTRODE ELECTRICAL RESISTIVITY

    SciTech Connect (OSTI)

    RUCKER DF; FINK JB; LOKE MH; MYERS DA

    2009-11-05T23:59:59.000Z

    Highly industrialized areas pose significant challenges for surface based electrical resistivity characterization and monitoring due to the high degree of metallic infrastructure. The infrastructure is typically several orders of magnitude more conductive than the desired targets, preventing the geophysicist from obtaining a clear picture of the subsurface. These challenges may be minimized if steel-cased wells are used as long electrodes. We demonstrate a method of using long electrodes in a complex nuclear waste facility to monitor a simulated leak from an underground storage tank. The leak was simulated by injecting high conductivity fluid in a perforated well and the resistivity measurements were made before and after the leak test. The data were processed in four dimensions, where a regularization procedure was applied in both the time and space domains. The results showed a lowered resistivity feature develop south of the injection site. The time lapsed regularization parameter had a strong influence on the differences in inverted resistivity between the pre and post datasets, potentially making calibration of the results to specific hydrogeologic parameters difficult.

  16. Determination of crack morphology parameters from service failures for leak-rate analyses

    SciTech Connect (OSTI)

    Wilkowski, G.; Ghadiali, N.; Paul, D. [Battelle Memorial Institute, Columbus, OH (United States)] [and others

    1997-04-01T23:59:59.000Z

    In leak-rate analyses described in the literature, the crack morphology parameters are typically not well agreed upon by different investigators. This paper presents results on a review of crack morphology parameters determined from examination of service induced cracks. Service induced cracks were found to have a much more tortuous flow path than laboratory induced cracks due to crack branching associated with the service induced cracks. Several new parameters such as local and global surface roughnesses, as well as local and global number of turns were identified. The effect of each of these parameters are dependent on the crack-opening displacement. Additionally, the crack path is typically assumed to be straight through the pipe thickness, but the service data show that the flow path can be longer due to the crack following a fusion line, and/or the number of turns, where the number of turns in the past were included as a pressure drop term due to the turns, but not the longer flow path length. These parameters were statistically evaluated for fatigue cracks in air, corrosion-fatigue, IGSCC, and thermal fatigue cracks. A refined version of the SQUIRT leak-rate code was developed to account for these variables. Sample calculations are provided in this paper that show how the crack size can vary for a given leak rate and the statistical variation of the crack morphology parameters.

  17. Aerosol penetration of leak pathways : an examination of the available data and models.

    SciTech Connect (OSTI)

    Powers, Dana Auburn

    2009-04-01T23:59:59.000Z

    Data and models of aerosol particle deposition in leak pathways are described. Pathways considered include capillaries, orifices, slots and cracks in concrete. The Morewitz-Vaughan criterion for aerosol plugging of leak pathways is shown to be applicable only to a limited range of particle settling velocities and Stokes numbers. More useful are sampling efficiency criteria defined by Davies and by Liu and Agarwal. Deposition of particles can be limited by bounce from surfaces defining leak pathways and by resuspension of particles deposited on these surfaces. A model of the probability of particle bounce is described. Resuspension of deposited particles can be triggered by changes in flow conditions, particle impact on deposits and by shock or vibration of the surfaces. This examination was performed as part of the review of the AP1000 Standard Combined License Technical Report, APP-GW-GLN-12, Revision 0, 'Offsite and Control Room Dose Changes' (TR-112) in support of the USNRC AP1000 Standard Combined License Pre-Application Review.

  18. Nowcasting Disaster Damage

    E-Print Network [OSTI]

    Kryvasheyeu, Yury; Obradovich, Nick; Moro, Esteban; Van Hentenryck, Pascal; Fowler, James; Cebrian, Manuel

    2015-01-01T23:59:59.000Z

    Could social media data aid in disaster response and damage assessment? Countries face both an increasing frequency and intensity of natural disasters due to climate change. And during such events, citizens are turning to social media platforms for disaster-related communication and information. Social media improves situational awareness, facilitates dissemination of emergency information, enables early warning systems, and helps coordinate relief efforts. Additionally, spatiotemporal distribution of disaster-related messages helps with real-time monitoring and assessment of the disaster itself. Here we present a multiscale analysis of Twitter activity before, during, and after Hurricane Sandy. We examine the online response of 50 metropolitan areas of the United States and find a strong relationship between proximity to Sandy's path and hurricane-related social media activity. We show that real and perceived threats -- together with the physical disaster effects -- are directly observable through the intens...

  19. EIGHTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE

    SciTech Connect (OSTI)

    Daugherty, W. L.

    2013-09-03T23:59:59.000Z

    A series of experiments to monitor the aging performance of Viton® GLT O-rings used in the Model 9975 package has been ongoing since 2004 at the Savannah River National Laboratory. Seventy tests using mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 ºF. They were leak-tested initially and have been tested periodically to determine if they meet the criterion of leak-tightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 ºF. High temperature aging continues for 23 GLT O-ring fixtures at 200 – 270 ºF. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 ºF and higher temperatures, and in 8 fixtures aging at 300 ºF. The remaining GLT O-ring fixtures aging at 300 ºF have been retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 ºF for 61 - 85 months, which is still bounding to O-ring temperatures during storage in KArea Complex (KAC). Based on expectations that the fixtures aging at 200 ºF will remain leaktight for a significant period yet to come, 2 additional fixtures began aging in 2011 at an intermediate temperature of 270 ºF, with hopes that they may reach a failure condition before the 200 ºF fixtures. High temperature aging continues for 6 GLT-S O-ring fixtures at 200 – 300 ºF. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 ºF. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 - 300 ºF for 41 - 45 months. Aging and periodic leak testing will continue for the remaining PCV fixtures.

  20. 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.

  1. NINTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE

    SciTech Connect (OSTI)

    Daugherty, W.

    2014-08-06T23:59:59.000Z

    A series of experiments to monitor the aging performance of Viton® GLT O-rings used in the Model 9975 package has been ongoing since 2004 at the Savannah River National Laboratory. One approach has been to periodically evaluate the leak performance of O-rings being aged in mock-up 9975 Primary Containment Vessels (PCVs) at elevated temperatures. Other methods such as compression-stress relaxation (CSR) tests and field surveillance are also on-going to evaluate O-ring behavior. Seventy tests using PCV mock-ups were assembled and heated to temperatures ranging from 200 to 450 ºF. They were leak-tested initially and have been tested periodically to determine if they continue to meet the leak-tightness criterion defined in ANSI standard N14.5-97. Due to material substitution, fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 ºF. High temperature aging continues for 23 GLT O-ring fixtures at 200 – 270 ºF. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 ºF and higher temperatures, and in 8 fixtures aging at 300 ºF. The earliest 300 °F GLT O-ring fixture failure was observed at 34 months. The remaining GLT O-ring fixtures aging at 300 ºF have been retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 ºF for 72 - 96 months, which bounds O-ring temperatures anticipated during storage in K-Area Complex (KAC). Based on expectations that the 200 ºF fixtures will remain leak-tight for a significant period yet to come, 2 additional fixtures began aging in 2011 at 270 ºF, with hopes that they may reach a failure condition before the 200 ºF fixtures, thus providing additional time to failure data. High temperature aging continues for 6 GLT-S O-ring fixtures at 200 – 300 ºF. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 ºF. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 - 300 ºF for 54 - 57 months. No additional O-ring failures have been observed since the last interim report was issued. Aging and periodic leak testing will continue for the remaining PCV fixtures. Additional irradiation of several fixtures is recommended to maintain a balance between thermal and radiation exposures similar to that experienced in storage, and to show the degree of consistency of radiation response between GLT and GLT-S O-rings.

  2. 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

  3. Shock Initiation of Damaged Explosives

    SciTech Connect (OSTI)

    Chidester, S K; Vandersall, K S; Tarver, C M

    2009-10-22T23:59:59.000Z

    Explosive and propellant charges are subjected to various mechanical and thermal insults that can increase their sensitivity over the course of their lifetimes. To quantify this effect, shock initiation experiments were performed on mechanically and thermally damaged LX-04 (85% HMX, 15% Viton by weight) and PBX 9502 (95% TATB, 5% Kel-F by weight) to obtain in-situ manganin pressure gauge data and run distances to detonation at various shock pressures. We report the behavior of the HMX-based explosive LX-04 that was damaged mechanically by applying a compressive load of 600 psi for 20,000 cycles, thus creating many small narrow cracks, or by cutting wedge shaped parts that were then loosely reassembled, thus creating a few large cracks. The thermally damaged LX-04 charges were heated to 190 C for long enough for the beta to delta solid - solid phase transition to occur, and then cooled to ambient temperature. Mechanically damaged LX-04 exhibited only slightly increased shock sensitivity, while thermally damaged LX-04 was much more shock sensitive. Similarly, the insensitive explosive PBX 9502 was mechanically damaged using the same two techniques. Since PBX 9502 does not undergo a solid - solid phase transition but does undergo irreversible or 'rachet' growth when thermally cycled, thermal damage to PBX 9502 was induced by this procedure. As for LX-04, the thermally damaged PBX 9502 demonstrated a greater shock sensitivity than mechanically damaged PBX 9502. The Ignition and Growth reactive flow model calculated the increased sensitivities by igniting more damaged LX-04 and PBX 9502 near the shock front based on the measured densities (porosities) of the damaged charges.

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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...

  9. 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...

  10. 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.

  11. 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...

  12. 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...

  13. 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...

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. 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,...

  19. 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....

  20. 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...

  1. 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...

  2. 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...

  3. 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,...

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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...

  9. 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...

  10. 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...

  11. 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...

  12. 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...

  13. 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...

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. Lower head creep rupture failure analysis associated with alternative accident sequences of the Three Mile Island Unit 2

    SciTech Connect (OSTI)

    Sang Lung, Chan [Swiss Federal Institute of Technology Zurich and Swiss Federal Nuclear Safety Inspectorate, Zurich, Switzerland, 8001 (Switzerland)

    2004-07-01T23:59:59.000Z

    The objective of this lower head creep rupture analysis is to assess the current version of MELCOR 1.8.5-RG against SCDAP/RELAP5 MOD 3.3kz. The purpose of this assessment is to investigate the current MELCOR in-vessel core damage progression phenomena including the model for the formation of a molten pool. The model for stratified molten pool natural heat transfer will be included in the next MELCOR release. Presently, MELCOR excludes the gap heat-transfer model for the cooling associated with the narrow gap between the debris and the lower head vessel wall. All these phenomenological models are already treated in SCDAP/RELAP5 using the COUPLE code to model the heat transfer of the relocated debris with the lower head based on a two-dimensional finite-element-method. The assessment should determine if current MELCOR capabilities adequately cover core degradation phenomena appropriate for the consolidated MELCOR code. Inclusion of these features should bring MELCOR much closer to a state of parity with SCDAP/RELAP5 and is a currently underway element in the MELCOR code consolidation effort. This assessment deals with the following analysis of the Three Mile Island Unit 2 (TMI-2) alternative accident sequences. The TMI-2 alternative accident sequence-1 includes the continuation of the base case of the TMI-2 accident with the Reactor Coolant Pumps (RCP) tripped, and the High Pressure Injection System (HPIS) throttled after approximately 6000 s accident time, while in the TMI-2 alternative accident sequence-2, the reactor coolant pumps is tripped after 6000 s and the HPIS is activated after 12,012 s. The lower head temperature distributions calculated with SCDAP/RELAP5 are visualized and animated with open source visualization freeware 'OpenDX'. (author)

  20. DESCRIPTION OF ACCIDENT MSU DRIVERS SIGNATURE

    E-Print Network [OSTI]

    Dyer, Bill

    of Commercial Policy Number) Motor vehicles that are owned, rented, leased, or loaned to Montana State's Name: MSU VEHICLE (VEHICLE #1) Issued Citation: YES NO Explain: Department: Phone: 994 - Vehicle Owner: Use of Vehicle: Vehicle: Make Model Year VIN: Plate Number: State: Description of Damage: Safety

  1. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect (OSTI)

    NONE

    1998-03-01T23:59:59.000Z

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  2. TACIS 91: Application of leak-before-break concept in VVER 440-230

    SciTech Connect (OSTI)

    Bartholome, G.; Faidy, C.; Franco, C. [and others

    1997-04-01T23:59:59.000Z

    The applicability of the leak-before-break (LBB) concept for primary piping in the first generation of WWER type plants in Russia is investigated. The procedures for LBB behavior used in France and Germany are applied, and the evaluation is discussed within the framework of the European Technical Assistance for the Community of Independent States (TACIS) project. Emphasis is placed on experimental validation of national and international engineering practice for evaluating and optimizing existing installations. Design criteria of WWER plants are compared to western standard design.

  3. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    SciTech Connect (OSTI)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01T23:59:59.000Z

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  4. The leak resistance of 2-inch N-80 API treaded tubular connection

    E-Print Network [OSTI]

    Weiner, Peter Douglas

    1961-01-01T23:59:59.000Z

    -UPS OF 2-INCH N-80 EUE TUBING 15 17 13 FRONTAL VIEW OF TEST TANK 14 TEST TA1K 15 PRESSURE TEST DATA SHEET 16 LONG DURATION TANK 18 19 21 22 THE LEAK RESISTANCE OF 2-INCH N-80 API ~ED TUBULAR CONNECTION INTROI3UCTION In recent years, well depths... tension until an equivalent pull of 18, 000 feet of tubing was exerted. on the tubing. Each specimen was subJected to from 50 to 100 thermocycles to simulate the shut-in and. flow conditions in an oil well and to increase the severity of the pressure...

  5. The concepts of leak before break and absolute reliability of NPP equipment and piping

    SciTech Connect (OSTI)

    Getman, A.F.; Komarov, O.V.; Sokov, L.M. [and others

    1997-04-01T23:59:59.000Z

    This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described.

  6. Assessments of fluid friction factors for use in leak rate calculations

    SciTech Connect (OSTI)

    Chivers, T.C. [Berkeley Technology Centre, Glos (United Kingdom)

    1997-04-01T23:59:59.000Z

    Leak before Break procedures require estimates of leakage, and these in turn need fluid friction to be assessed. In this paper available data on flow rates through idealized and real crack geometries are reviewed in terms of a single friction factor k It is shown that for {lambda} < 1 flow rates can be bounded using correlations in terms of surface R{sub a} values. For {lambda} > 1 the database is less precise, but {lambda} {approx} 4 is an upper bound, hence in this region flow calculations can be assessed using 1 < {lambda} < 4.

  7. 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.

  8. 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.

  9. 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

  10. 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,...

  11. 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,

  12. 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

  13. 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

  14. 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...

  15. 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.

  16. 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.

  17. 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.

  18. 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...

  19. 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

  20. 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...

  1. 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...

  2. 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...

  3. 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...

  4. 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

  5. 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...

  6. 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...

  7. 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...

  8. 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

  9. 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...

  10. 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...

  11. 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...

  12. 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

  13. 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

  14. 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 ...

  15. 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...

  16. 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

  17. 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

  18. 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

  19. 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...

  20. 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

  1. 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...

  2. 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

  3. 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...

  4. 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...

  5. 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

  6. 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...

  7. 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...

  8. 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

  9. 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...

  10. 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...

  11. 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...

  12. Methods for detecting and locating leaks in containment facilities using electrical potential data and electrical resistance tomographic imaging techniques

    DOE Patents [OSTI]

    Daily, William D. (Livermore, CA); Laine, Daren L. (San Anotonio, TX); Laine, Edwin F. (Penn Valley, CA)

    2001-01-01T23:59:59.000Z

    Methods are provided for detecting and locating leaks in liners used as barriers in the construction of landfills, surface impoundments, water reservoirs, tanks, and the like. Electrodes are placed in the ground around the periphery of the facility, in the leak detection zone located between two liners if present, and/or within the containment facility. Electrical resistivity data is collected using these electrodes. This data is used to map the electrical resistivity distribution beneath the containment liner or between two liners in a double-lined facility. In an alternative embodiment, an electrode placed within the lined facility is driven to an electrical potential with respect to another electrode placed at a distance from the lined facility (mise-a-la-masse). Voltage differences are then measured between various combinations of additional electrodes placed in the soil on the periphery of the facility, the leak detection zone, or within the facility. A leak of liquid through the liner material will result in an electrical potential distribution that can be measured at the electrodes. The leak position is located by determining the coordinates of an electrical current source pole that best fits the measured potentials with the constraints of the known or assumed resistivity distribution.

  13. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    SciTech Connect (OSTI)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    1989-08-01T23:59:59.000Z

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

  14. 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.

  15. Damage and Damage Prediction for Wood Shearwalls Subjected to Simulated Earthquake Loads

    E-Print Network [OSTI]

    Gupta, Rakesh

    Damage and Damage Prediction for Wood Shearwalls Subjected to Simulated Earthquake Loads John W damaged resulting in large financial losses. Societal demands for damage-limiting design philosophies and better predict damage to woodframe structures. This paper examines damage to the lateral load carrying

  16. 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

  17. Excitation optimization for damage detection

    SciTech Connect (OSTI)

    Bement, Matthew T [Los Alamos National Laboratory; Bewley, Thomas R [UCSD

    2009-01-01T23:59:59.000Z

    A technique is developed to answer the important question: 'Given limited system response measurements and ever-present physical limits on the level of excitation, what excitation should be provided to a system to make damage most detectable?' Specifically, a method is presented for optimizing excitations that maximize the sensitivity of output measurements to perturbations in damage-related parameters estimated with an extended Kalman filter. This optimization is carried out in a computationally efficient manner using adjoint-based optimization and causes the innovations term in the extended Kalman filter to be larger in the presence of estimation errors, which leads to a better estimate of the damage-related parameters in question. The technique is demonstrated numerically on a nonlinear 2 DOF system, where a significant improvement in the damage-related parameter estimation is observed.

  18. Location of Leaks in Pressure Testable Direct Burial Steam Distribution Conduits 

    E-Print Network [OSTI]

    Sittel, M. G.; Messock, R. K.

    1993-01-01T23:59:59.000Z

    , resulting in increased thermal energy losses and eventual damage to the steam line. Breaches in the outer conduit are difficult to locate, and damage to the steam line may progress until the entire line requires replacement. Thermal energy losses are high...

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. DAMAGE LOCALIZATION USING LOAD VECTORS Dionisio Bernal

    E-Print Network [OSTI]

    Bernal, Dionisio

    DAMAGE LOCALIZATION USING LOAD VECTORS Dionisio Bernal Associate Professor Department of Civil: A technique to localize damage in structures that can be treated as linear in the pre and post-damage state is presented. Central to the approach is the computation of a set of vectors, designated as Damage Locating

  4. Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. [Brookhaven National Lab., Upton, NY (United States); Siu, N. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1991-12-31T23:59:59.000Z

    This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff`s follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

  5. Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. (Brookhaven National Lab., Upton, NY (United States)); Siu, N. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

    1991-01-01T23:59:59.000Z

    This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff's follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

  6. 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...

  7. 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

  8. 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

  9. 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

  10. 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

  11. Specialist meeting on leak before break in reactor piping and vessels

    SciTech Connect (OSTI)

    Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

    1997-04-01T23:59:59.000Z

    A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

  12. Practical applications of the R6 leak-before-break procedure

    SciTech Connect (OSTI)

    Bouchard, P.J.

    1997-04-01T23:59:59.000Z

    A forthcoming revision to the R6 Leak-before-Break Assessment Procedure is briefly described. Practical application of the LbB concepts to safety-critical nuclear plant is illustrated by examples covering both low temperature and high temperature (>450{degrees}C) operating regimes. The examples highlight a number of issues which can make the development of a satisfactory LbB case problematic: for example, coping with highly loaded components, methodology assumptions and the definition of margins, the effect of crack closure owing to weld residual stresses, complex thermal stress fields or primary bending fields, the treatment of locally high stresses at crack intersections with free surfaces, the choice of local limit load solution when predicting ligament breakthrough, and the scope of calculations required to support even a simplified LbB case for high temperature steam pipe-work systems.

  13. Crack shape developments and leak rates for circumferential complex-cracked pipes

    SciTech Connect (OSTI)

    Brickstad, B.; Bergman, M. [SAQ Inspection Ltd., Stockholm (Sweden)

    1997-04-01T23:59:59.000Z

    A computerized procedure has been developed that predicts the growth of an initial circumferential surface crack through a pipe and further on to failure. The crack growth mechanism can either be fatigue or stress corrosion. Consideration is taken to complex crack shapes and for the through-wall cracks, crack opening areas and leak rates are also calculated. The procedure is based on a large number of three-dimensional finite element calculations of cracked pipes. The results from these calculations are stored in a database from which the PC-program, denoted LBBPIPE, reads all necessary information. In this paper, a sensitivity analysis is presented for cracked pipes subjected to both stress corrosion and vibration fatigue.

  14. 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 ...

  15. 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,...

  16. 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...

  17. 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...

  18. 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...

  19. 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 ...

  20. 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 ...