National Library of Energy BETA

Sample records for leaks damage accidents

  1. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    SciTech Connect (OSTI)

    Sdouz, Gert [ARC Seibersdorf Research GmbH, Viktor Kaplan-Strasse 2, 2700 Wr. Neustadt (Austria)

    2006-07-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a 'Large Break LOCA'-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a 'Station Blackout'-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors. (authors)0.

  2. 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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  3. Calculation notes for surface leak resulting in pool, TWRS FSAR accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    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. Calculation Notes for Subsurface Leak Resulting in Pool, TWRS FSAR Accident Analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    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.

  5. Calculation notes in support of TWRS FSAR spray leak accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document contains the detailed calculations that support the spray leak accident analysis in the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR). The consequence analyses in this document form the basis for the selection of controls to mitigate or prevent spray leaks throughout TWRS. Pressurized spray leaks can occur due to a breach in containment barriers along transfer routes, during waste transfers. Spray leaks are of particular safety concern because, depending on leak dimensions, and waste pressure, they can be relatively efficient generators of dispersible sized aerosols that can transport downwind to onsite and offsite receptors. Waste is transferred between storage tanks and between processing facilities and storage tanks in TWRS through a system of buried transfer lines. Pumps for transferring waste and jumpers and valves for rerouting waste are located inside below grade pits and structures that are normally covered. Pressurized spray leaks can emanate to the atmosphere due to breaches in waste transfer associated equipment inside these structures should the structures be uncovered at the time of the leak. Pressurized spray leaks can develop through holes or cracks in transfer piping, valve bodies or pump casings caused by such mechanisms as corrosion, erosion, thermal stress, or water hammer. Leaks through degraded valve packing, jumper gaskets, or pump seals can also result in pressurized spray releases. Mechanisms that can degrade seals, packing and gaskets include aging, radiation hardening, thermal stress, etc. An1782other common cause for spray leaks inside transfer enclosures are misaligned jumpers caused by human error. A spray leak inside a DST valve pit during a transfer of aging waste was selected as the bounding, representative accident for detailed analysis. Sections 2 through 5 below develop this representative accident using the DOE- STD-3009 format. Sections 2 describes the unmitigated and mitigated accident scenarios evaluated to determine the need for safety class SSCs or TSR controls. Section 3 develops the source terms associated with the unmitigated and mitigated accident scenarios. Section 4 estimates the radiological and toxicological consequences for the unmitigated and mitigated scenarios. Section 5 compares the radiological and toxicological consequences against the TWRS evaluation guidelines. Section 6 extrapolates from the representative accident case to other represented spray leak sites to assess the conservatism in using the representative case to define controls for other postulated spray leak sites throughout TWRS. Section 7 discusses the sensitivities of the consequence analyses to the key parameters and assumptions used in the analyses. Conclusions are drawn in Section 8. The analyses herein pertain to spray leaks initiated due to internal mechanisms (e.g., corrosion, erosion, thermal stress, etc). External initiators of spray leaks (e.g., excavation accidents), and natural phenomena initiators (e.g., seismic events) are to be covered in separate accident analyses.

  6. Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents.

    SciTech Connect (OSTI)

    Majumdar, S.; Diercks, D. R.; Shack, W. J.; Energy Technology

    2002-05-01

    This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents.

  7. 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)

    1994-04-01

    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.

  8. Leak detection/verification

    SciTech Connect (OSTI)

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

    1997-04-01

    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.

  9. Precursors to potential severe core damage accidents, 1986: A status report: Main report and Appendixes A,B, and C

    SciTech Connect (OSTI)

    Minarick, J W; Harris, J D; Austin, P N; Cletcher, J W; Hagen, E W

    1988-05-01

    The Accident Sequence Precursor Program reviews licensee event reports of operational events that have occurred at LWRs to identify and categorize precursors to potential severe core-damage accidents. Accident sequences considered in the study are those associated with inadequate core cooling. Accident sequence precursors are events that are important elements in such sequences. Such precursors could be infrequent initiating events or equipment failures that, when coupled with one or more postulated events, could result in a plant condition with inadequate core cooling. Originally proposed in the Risk Assessment Review Group Report (Lewis Committee report) in 1978, the study - subsequently named the Accident Sequence Precursor Program - was initiated at the Nuclear Operations Analysis Center in 1979. Earlier reports by the program involved assessment of events that occurred in 1969-1981 and 1984-1985. The present report involves the assessment of events that occurred during 1986. A nuclear plant has safety systems for mitigating the consequences of accidents or off-normal initiating events that may occur during the course of plant operation. These systems are built to high-quality standards and are redundant; nonetheless, they have a nonzero probability of failing or being in a failed state when required to operate. This report uses LERs and other plant data, estimated system unavailabilities, the expected average frequency of initiating events (LOFWs, LOOPs, LOCAs), and event details to evaluate the potential impact of the following two situations.

  10. Reducing Your Leak Rate Without Repairing Leaks 

    E-Print Network [OSTI]

    Beals, C.

    2005-01-01

    below the header pressure have the added advantage of reducing the air consumption of equipment, as well as reducing the leak rate. Turn Off the Air to Idle Equipment In most plants, when production equipment operators shut off their equipment... of the reason why plant personnel often find leak repair unproductive. The other reason relates to the compressor controls. If the plant had several lubricated rotary screw compressors operating in modulation, repairing 2,000 cfm in leaks may only reduce...

  11. Chlorofluorocarbon leak detection technology

    SciTech Connect (OSTI)

    Munday, E.B.

    1990-12-01

    There are about 590 large coolant systems located at the Portsmouth Gaseous Diffusion Plant (PORTS) and the Paducah Gaseous Diffusion Plant (PGDP) leaking nearly 800,000 lb of R-114 refrigerant annually (1989 estimate). A program is now under way to reduce the leakage to 325,000 lb/year -- an average loss of 551 lb/year (0.063 lb/h) per coolant system, some of which are as large as 800 ft. This report investigates leak detection technologies that can be used to locate leaks in the coolant systems. Included are descriptions, minimum leak detection rate levels, advantages, disadvantages, and vendor information on the following technologies: bubbling solutions; colorimetric leak testing; dyes; halogen leak detectors (coronea discharge detectors; halide torch detectors, and heated anode detectors); laser imaging; mass spectroscopy; organic vapor analyzers; odorants; pressure decay methods; solid-state electrolytic-cell gas sensors; thermal conductivity leak detectors; and ultrasonic leak detectors.

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

    E-Print Network [OSTI]

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

    2014-01-01

    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 ...

  13. 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

  14. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    2000-01-10

    Four bounding accidents postulated 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 a hydrogen explosion, and a fire breaching filter vessel and enclosure. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  15. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    RITTMANN, P.D.

    1999-10-07

    Three bounding accidents postdated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing, and a hydrogen explosion. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  16. Leak detection aid

    DOE Patents [OSTI]

    Steeper, T.J.

    1989-12-26

    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.

  17. Leak detection aid

    DOE Patents [OSTI]

    Steeper, Timothy J. (Graniteville, SC)

    1989-01-01

    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.

  18. 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

    1994-06-01

    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.

  19. Gaseous leak detector

    DOE Patents [OSTI]

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

    1988-01-01

    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.

  20. Improved gaseous leak detector

    DOE Patents [OSTI]

    Juravic, F.E. Jr.

    1983-10-06

    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.

  1. Sensitive hydrogen leak detector

    DOE Patents [OSTI]

    Myneni, Ganapati Rao (Yorktown, VA)

    1999-01-01

    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.

  2. Sensitive hydrogen leak detector

    DOE Patents [OSTI]

    Myneni, G.R.

    1999-08-03

    A sensitive hydrogen leak detector system is described which uses 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. 1 fig.

  3. 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)

    1996-01-01

    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.

  4. 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)

    2008-05-20

    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.

  5. 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)

    1994-08-01

    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.

  6. Leak test fitting

    DOE Patents [OSTI]

    Pickett, Patrick T. (Kettering, OH)

    1981-01-01

    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.

  7. Leaking Pipelines: Doctoral Student Family Formation

    E-Print Network [OSTI]

    Serrano, Christyna M.

    2008-01-01

    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

  8. Accident management for severe accidents

    SciTech Connect (OSTI)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs.

  9. AndroidLeaks: Automatically Detecting Potential Privacy Leaks In Android Applications

    E-Print Network [OSTI]

    Chen, Hao

    AndroidLeaks: Automatically Detecting Potential Privacy Leaks In Android Applications on a Large of sensitive information, they may leak it carelessly or maliciously. Google's Android operating systemLeaks, a static analysis framework for automatically finding poten- tial leaks of sensitive information in Android

  10. Accident Investigations

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

    1996-04-26

    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

  11. Accident Investigations

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

    1995-10-26

    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

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

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

    3: Distribution System Audits, Leak Detection, and Repair Best Management Practice 3: Distribution System Audits, Leak Detection, and Repair A distribution system audit, leak...

  13. Accident Investigations

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

    2011-03-04

    This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities. Supersedes DOE O 225.1A. Cancels DOE G 225.1A-1.

  14. Stochastic Consequence Analysis for Waste Leaks

    SciTech Connect (OSTI)

    HEY, B.E.

    2000-05-31

    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

  15. Investigating leaking underground storage tanks 

    E-Print Network [OSTI]

    Upton, David Thompson

    1989-01-01

    general methodology for many geologic regions where stratigraphic and hydrogeologic conditions are likely to be similar. Ultimately, the goal of any investigator or owner is to obtain the necessary information in order to satisfy the concerns... INVESTIGATING LEAKING UNDERGROUND STORAGE TANKS A Thesis by DAVID THOMPSON UPTON Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE August 1989...

  16. Improved dose assessment in a nuclear reactor accident using the old and new ICRP methodologies 

    E-Print Network [OSTI]

    Yoon, Suk-Chul

    1987-01-01

    using WASH-1400 methodologies will be presented later with discussion of the previous work. THEORY AND MODELING Atmospheric Dispersion Radionuclides leaking from the containment during a severe nuclear accident are dispersed continually... generic set of accident releases that could provide useful insights into improved safety action. A set of five groups of source terms were proposed (BI82) to encompass the full spectrum of severe accident release possibilities, these were called Siting...

  17. Vacuum leak detector and method

    DOE Patents [OSTI]

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

    1983-01-01

    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.

  18. Leak detection on an ethylene pipeline

    SciTech Connect (OSTI)

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

    1995-12-31

    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.

  19. High sensitivity leak detection method and apparatus

    DOE Patents [OSTI]

    Myneni, G.R.

    1994-09-06

    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.

  20. High sensitivity leak detection method and apparatus

    DOE Patents [OSTI]

    Myneni, Ganapatic R. (Grafton, VA)

    1994-01-01

    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.

  1. Developing a knowledge base for the management of severe accidents

    SciTech Connect (OSTI)

    Nelson, W.R.; Jenkins, J.P.

    1986-01-01

    Prior to the accident at Three Mile Island, little attention was given to the development of procedures for the management of severe accidents, that is, accidents in which the reactor core is damaged. Since TMI, however, significant effort has been devoted to developing strategies for severe accident management. At the same time, the potential application of artificial intelligence techniques, particularly expert systems, to complex decision-making tasks such as accident diagnosis and response has received considerable attention. The need to develop strategies for accident management suggests that a computerized knowledge base such as used by an expert system could be developed to collect and organize knowledge for severe accident management. This paper suggests a general method which could be used to develop such a knowledge base, and how it could be used to enhance accident management capabilities.

  2. Leak checker data logging system

    DOE Patents [OSTI]

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

    1996-09-03

    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.

  3. Leak checker data logging system

    DOE Patents [OSTI]

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

    1996-01-01

    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.

  4. Accident progression event tree analysis for postulated severe accidents at N Reactor

    SciTech Connect (OSTI)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

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

    SciTech Connect (OSTI)

    Ru Chen; Ian Kaye

    2012-03-12

    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.

  6. Geomechanical analysis to predict the oil leak at the wellbores in Big Hill Strategic Petroleum Reserve

    SciTech Connect (OSTI)

    Park, Byoung Yoon

    2014-02-01

    Oil leaks were found in wellbores of Caverns 105 and 109 at the Big Hill Strategic Petroleum Reserve site. According to the field observations, two instances of casing damage occurred at the depth of the interbed between the caprock bottom and salt top. A three dimensional finite element model, which contains wellbore element blocks and allows each cavern to be configured individually, is constructed to investigate the wellbore damage mechanism. The model also contains element blocks to represent interface between each lithology and a shear zone to examine the interbed behavior in a realistic manner. The causes of the damaged casing segments are a result of vertical and horizontal movements of the interbed between the caprock and salt dome. The salt top subsides because the volume of caverns below the salt top decrease with time due to salt creep closure, while the caprock subsides at a slower rate because the caprock is thick and stiffer. This discrepancy yields a deformation of the well. The deformed wellbore may fail at some time. An oil leak occurs when the wellbore fails. A possible oil leak date of each well is determined using the equivalent plastic strain failure criterion. A well grading system for a remediation plan is developed based on the predicted leak dates of each wellbore.

  7. Type B Accident Investigation Board Report Grout Injection Operator...

    Energy Savers [EERE]

    and no damage to any structures inside the calvareum (i.e., no evidence of brain injury). Page 16 2.4. Investigation Readiness and Accident Scene Preservation The...

  8. Accident information needs

    SciTech Connect (OSTI)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  9. Accident information needs

    SciTech Connect (OSTI)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

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

    SciTech Connect (OSTI)

    VENETZ TJ; WASHENFELDER D; JOHNSON J; GIRARDOT C

    2012-01-25

    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.

  11. 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 ...

  12. Fuel performance during severe accidents. [PWR

    SciTech Connect (OSTI)

    Buescher, B.J.; Gruen, G.E.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. This program is underway in the Power Burst Facility at the Idaho National Engineering Laboratory. In preparation for the first test, predictions have been performed using the TRAC-BD1 computer. This paper presents the calculated results showing a slow heatup to 2400 K over 5 hours, and the analysis includes accelerated oxidation of the zirconium cladding at temperatures above 1850 K.

  13. New system pinpoints leaks in ethylene pipeline

    SciTech Connect (OSTI)

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

    1995-04-01

    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.

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

    SciTech Connect (OSTI)

    Salay, Michael; Kalinich, Donald A.; Gauntt, Randall O.; Radel, Tracy E.

    2008-10-01

    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. Preliminary analysis of tank 241-C-106 dryout due to large postulated leak and vaporization

    SciTech Connect (OSTI)

    Piepho, M.G.

    1994-12-01

    This analysis assumes that there is a hypothetical large leak at the bottom of Tank 241-C-106 which initiates the dryout of the tank. The time required for a tank to dryout after a leak is of interest for safety reasons. As a tank dries out, its temperature is expected to increase which could affect the structural integrity of the concrete tank dome. Hence, it is of interest to know how fast and how high the temperature in a leaky tank increases, so that mitigation procedures can be planned and implemented in a timely manner. This analysis is focused on tank 241-C-106, which is known to be high thermal tank. The objective of the study was to determine how long it would take for tank 241-C-106 to reach 350 degrees Fahrenheit (about 177 degrees Centigrade) after a postulated large leak develops at the bottom center of the tank. The temperature of 350 degrees Fahrenheit is the minimum temperature that can cause structural damage to concrete (ACI 1992). The postulated leak at the bottom of the tank and the resulting dryout of the sludge in the tank make this analysis different from previous thermal analyses of the C-106 tank and other tanks, especially the double-shell tanks which are mostly liquid.

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

    SciTech Connect (OSTI)

    Washenfelder, Dennis J.

    2014-04-10

    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.

  17. 1999 Leak Detection and Monitoring and Mitigation Strategy Update

    SciTech Connect (OSTI)

    OHL, P.C.

    1999-09-23

    This document is a complete revision of WHC-SD-WM-ES-378, Rev 1. This update includes recent developments in Leak Detection, Leak Monitoring, and Leak Mitigation technologies, as well as, recent developments in single-shell tank retrieval technologies. In addition, a single-shell tank retrieval release protection strategy is presented.

  18. 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)

    2009-02-03

    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.

  19. Severe accident progression perspectives based on IPE results

    SciTech Connect (OSTI)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T.; Drouin, M.

    1996-08-01

    Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here.

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

    SciTech Connect (OSTI)

    Clinton, J.H.

    1994-09-01

    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.

  1. Management of severe accidents

    SciTech Connect (OSTI)

    Jankowski, M.W.

    1988-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery management concentrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk, and goes further in considering and formulating the key issue: The most fruitful path to follow in reducing risk even further is through the planning of accident management.

  2. Management of vacuum leak-detection processes, standards, and calibration

    SciTech Connect (OSTI)

    Wilson, N.G.

    1984-01-01

    Vacuum leak detection requires integrated management action to ensure the successful production of apparatus having required leak tightness. Implementation of properly planned, scheduled, and engineering procedures and test arrangements are an absolute necessity to prevent unexpected, impractical, technically inadequate, or unnecessarily costly incidents in leak-testing operations. The use of standard procedures, leak standards appropriate to the task, and accurate calibration systems or devices is necessary to validate the integrity of any leak-test procedure. In this paper, the need for implementing these practices is discussed using case histories of typical examples of large complex vacuum systems. Aggressive management practices are of primary importance throughout a project's life cycle to ensure the lowest cost; this includes successful leak testing of components. It should be noted that the opinions and conclusions expressed in this paper are those of the author and are not those of the Los Alamos National Laboratory or the Department of Energy.

  3. Managing an Effective Leak Sealing Program 

    E-Print Network [OSTI]

    Rinz, W. H.

    1980-01-01

    of steam/day S/hr Rate of stearn leakage and cost for valves with spindle diameters 1/2", 1" and 2" Fig. 2 STEAM LOSS/ENERGY COST ESTIMATION OF TYPICAL VALVE SPINDLE PACKING LEAKS 389 3.00 4.00 5.00 6.001.00 2.00242016 .k, .4, k, .4... must establish occur when a forced shutdown of a process unit takes the direction and controls necessary to ensure a place. In addition to financial losses, leakage successful energy saving program. from any system may cause the potential dangers...

  4. Air pollutant penetration through airflow leaks into buildings

    E-Print Network [OSTI]

    Liu, De-Ling

    2002-01-01

    leaks in the building envelope was advanced by performingadvanced our knowledge, they have not fully elucidated the extent to which particles penetrate building envelopes.

  5. Estimation of Gas Leak Rates Through Very Small Orifices

    Office of Scientific and Technical Information (OSTI)

    Estimation of Gas Leak Rates Through Very Small Orifices and Channels by Herbert J. Bomelburg February 1977 Prepared for the Nuclear Regulatory Commission -..- Pacific Northwest...

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

    SciTech Connect (OSTI)

    Trapp, D.

    2014-08-25

    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.

  7. Human factors review for nuclear power plant severe accident sequence analysis

    SciTech Connect (OSTI)

    Krois, P.A.; Haas, P.M.

    1985-01-01

    The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release.

  8. Accident resistant transport container

    DOE Patents [OSTI]

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

    1980-01-01

    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.

  9. Accident motivates scholarship recipient

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

    accident, college wasn't on my mind. I was all about sports," said Leyba, who played football, basketball, and ran track at the small Northern New Mexico school. "I had no idea...

  10. Accident management information needs

    SciTech Connect (OSTI)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  11. 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.

    2012-09-10

    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.

  12. 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

    2010-03-01

    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.

  13. A new blowdown compensation scheme for boiler leak detection

    E-Print Network [OSTI]

    Marquez, Horacio J.

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

  14. 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)

    2014-01-20

    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.

  15. ADEL: An Automatic Detector of Energy Leaks for Smartphone Applications

    E-Print Network [OSTI]

    Mao, Zhuoqing Morley

    ADEL: An Automatic Detector of Energy Leaks for Smartphone Applications Lide Zhang Mark S. Gordon pdinda@northwestern.edu Google Inc. Mountain View, CA, USA leiyang@google.com ABSTRACT Energy leaks occur when applications use energy to perform use- less tasks, a surprisingly common occurrence

  16. Spills and leaks Associated with Shale Gas Development

    E-Print Network [OSTI]

    Walter, M.Todd

    1 Spills and leaks Associated with Shale Gas Development (Updated April 27th , 2012) Brief of toxic chemicals, contaminated water, or hazardous materials. Spills and leaks associated with shale gas associated with shale gas development will depend on the pace and scale with which the industry grows

  17. FORMALISM HELPS IN DESCRIBING ACCIDENTS Peter Ladkin, Universitt Bielefeld, Germany

    E-Print Network [OSTI]

    Ladkin, Peter B.

    , such reasoning engineering is both essential and non-trivial. Accident reports in aviation present careful disagreement systems resulting from maintenance-induced damage leading to the separation of the No. 1 engine maintenance procedures which led to failure of the pylon structure. We shall analyse this statement

  18. Pressure Change Measurement Leak Testing Errors

    SciTech Connect (OSTI)

    Pryor, Jeff M; Walker, William C

    2014-01-01

    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.

  19. 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)

    1997-04-01

    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.

  20. Commercial SNF Accident Release Fractions

    SciTech Connect (OSTI)

    J. Schulz

    2004-11-05

    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.

  1. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room News Publications Traditional Knowledge KiosksAbout UsAboutWeb PoliciesAccidentAccident

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

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01

    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.

  3. A comparative analysis of accident risks in fossil, hydro, and nuclear energy chains

    SciTech Connect (OSTI)

    Burgherr, P.; Hirschberg, S.

    2008-07-01

    This study presents a comparative assessment of severe accident risks in the energy sector, based on the historical experience of fossil (coal, oil, natural gas, and LPG (Liquefied Petroleum Gas)) and hydro chains contained in the comprehensive Energy-related Severe Accident Database (ENSAD), as well as Probabilistic Safety Assessment (PSA) for the nuclear chain. Full energy chains were considered because accidents can take place at every stage of the chain. Comparative analyses for the years 1969-2000 included a total of 1870 severe ({>=} 5 fatalities) accidents, amounting to 81,258 fatalities. Although 79.1% of all accidents and 88.9% of associated fatalities occurred in less developed, non-OECD countries, industrialized OECD countries dominated insured losses (78.0%), reflecting their substantially higher insurance density and stricter safety regulations. Aggregated indicators and frequency-consequence (F-N) curves showed that energy-related accident risks in non-OECD countries are distinctly higher than in OECD countries. Hydropower in non-OECD countries and upstream stages within fossil energy chains are most accident-prone. Expected fatality rates are lowest for Western hydropower and nuclear power plants; however, the maximum credible consequences can be very large. Total economic damages due to severe accidents are substantial, but small when compared with natural disasters. Similarly, external costs associated with severe accidents are generally much smaller than monetized damages caused by air pollution.

  4. Heat leak performance of SSC collider dipole magnets

    SciTech Connect (OSTI)

    Weisend, J.G. II; Levin, M.; Franks, D.; Pletzer, R.; Augustynowicz, S.; McInturff, A.D.; Boroski, W.B.

    1993-09-01

    The large number of superconducting dipoles in the SSC results in a stringent heat leak budget for each dipole. Ensuring that the dipoles meet this budget is vital to the successful operation or the collider. This work surveys heat leak measurements taken during 4 different magnet string tests. These tests involved both 40 mm and SO mm aperture dipoles. In these experiments the heat leak to the 80 K shield, 20 K shield and cold mass are measured. The results are compared to predictions from a computational thermal model of the dipole cryostat. Discrepancies are seen between the predicted and measured values. Possible explanations for these discrepancies are given.

  5. 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

  6. Statistical approaches to leak detection for geological sequestration

    E-Print Network [OSTI]

    Haidari, Arman S

    2011-01-01

    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 ...

  7. From MSU News Service Leaking cisterns, inoperable solar

    E-Print Network [OSTI]

    Maxwell, Bruce D.

    From MSU News Service Leaking cisterns, inoperable solar panels and a local populace that didn development skeleton littering the landscape of Kenya." Seven members of the MSU student chapter of Engineers

  8. Pinch valves fight clogging, leaking and wear in FGD systems

    SciTech Connect (OSTI)

    Schneider, L.

    1982-12-01

    Pinch valves can provide a non-sticking, non-leaking, low maintenance system capable of controlling the flow of abrasive limestone slurries such as are found in flue gas desulphurisation units.

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

    E-Print Network [OSTI]

    Choi, Changrak

    2012-01-01

    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 ...

  10. Air Leaks in Unexpected Places | Department of Energy

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

    National Renewable Energy Laboratory What does this mean for me? Don't waste money heating a room with air leaks, learn how to identify and seal them up tight One of the...

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

    E-Print Network [OSTI]

    Chatzigeorgiou, Dimitrios

    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 ...

  12. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect (OSTI)

    El-Genk, Mohamed

    2012-10-19

    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.

  13. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect (OSTI)

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

    2014-10-01

    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.

  14. Electrical shock accident investigation

    SciTech Connect (OSTI)

    Not Available

    1994-09-30

    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.

  15. 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,...

  16. ATWS at Browns Ferry Unit One - accident sequence analysis

    SciTech Connect (OSTI)

    Harrington, R.M.; Hodge, S.A.

    1984-07-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence the quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.

  17. Sandia Energy - Severe Accident Modeling

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

    nuclear energy efforts by developing risk margins, creating risk assessments, sequencing nuclear reactor accident progression, and performing reactor consequence modeling. Severe...

  18. Apparatus and method for detecting leaks in piping

    DOE Patents [OSTI]

    Trapp, D.J.

    1994-12-27

    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.

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

    DOE Patents [OSTI]

    Henning, Carl D. (Livermore, CA)

    1993-01-01

    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.

  20. Industry program needed for nuclear accident management

    SciTech Connect (OSTI)

    Klopp, G.T

    1989-05-01

    This paper addresses the need for a management program for nuclear power accidents. According to the author, the tools and technology for severe accident management exist. The need for a clear, realistic definition of nuclear accident program requirements is discussed.

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

    SciTech Connect (OSTI)

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

    1997-04-01

    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.

  2. First Responders and Criticality Accidents

    SciTech Connect (OSTI)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  3. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-12-31

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  4. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  5. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  6. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. (Oak Ridge National Lab., TN (United States))

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

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

    Office of Environmental Management (EM)

    Incident Reporting System Computerized Accident Incident Reporting System CAIRS Database The Computerized AccidentIncident Reporting System is a database used to collect and...

  8. Gas leak characteristics of inner packaging components used in the D0T-Spec 6M container

    SciTech Connect (OSTI)

    Taylor, J.M.

    1985-09-01

    A test program was conducted by Pacific Northwest Laboratory to determine the gas leak characteristics of metal food pack cans and 2R vessels used to package radioactive material in a D0T 6M specification container. It can be concluded from the tests performed that the inner packaging components (2R vessel, metal product cans) used with a 6M container can be sealed so that they will be gas tight (<10/sup -5/ cc/sec) under elevated temperature and pressure and impact conditions. To maintain gas tight seals under accident conditions, the metal cans must be sealed with a properly adjusted can-sealing machine; the threads of the 2R vessel must be luted with a sealing compound such as a silicone rubber compound; and the metal cans must be protected inside the 2R vessel with spacer plates and impact absorbers. 4 refs., 37 figs.

  9. Assessment of two BWR accident management strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs.

  10. 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)

    1997-04-01

    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.

  11. Methodology to quantify leaks in aerosol sampling system components 

    E-Print Network [OSTI]

    Vijayaraghavan, Vishnu Karthik

    2004-11-15

    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...

  12. The feasibility of electrophoretic repair of impoundment leaks 

    E-Print Network [OSTI]

    Han, Ji-Seok

    2002-01-01

    to the impoundment. The cathode is placed inside and the anode is placed outside the impoundment. An electric field is imposed externally across the leaks of geomembrane liner. The negative charged clay particles are repelled by the cathode and attracted by the anode...

  13. 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

  14. DEVELOPMENT OF A METHODOLOGY TO PREDICT AND PREVENT LEAKS CAUSED

    E-Print Network [OSTI]

    Beckermann, Christoph

    to produce steel castings that are free from macroporosity (i.e., shrinkage porosity large enoughDEVELOPMENT OF A METHODOLOGY TO PREDICT AND PREVENT LEAKS CAUSED BY MICROPOROSITY IN STEEL CASTINGS to be detectable by radiographic testing). No risering rules currently exist to produce castings free from

  15. AIR SEALING Seal air leaks and save energy!

    E-Print Network [OSTI]

    Oak Ridge National Laboratory

    Kitchen Range Hood Kitchen and bath vents provide spot ventilation Annual Energy Costs for 1300 sq. ft AND RENEWABLE ENERGY · U.S. DEPARTMENT OF ENERGY #12; W H A T A RAIR 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

  16. INFORMAL REPORT DETECTION OF INTERSTATE LIQUIDS PIPELINE LEAKS

    E-Print Network [OSTI]

    with Battelle Memorial Institute and the Colonial Pipeline Company #12;ABSTRACT The approximately 200,000-mile half of the crude oil and petroleum products (gasoline, kerosene, home heating oils, diesel fuels be employed by pipeline companies would be the early detection of leaks while they are still small, that is

  17. Enhanced detection of groundwater contamination from a leaking waste disposal site by microbial community profiles

    E-Print Network [OSTI]

    Vermont, University of

    Enhanced detection of groundwater contamination from a leaking waste disposal site by microbial into the subsurface from leaking landfills. Detecting leachate contamination using statistical techniques of groundwater contamination. We sampled profiles of the microbial community from monitoring wells surrounding

  18. Accuracy of Distributed Optical Fiber Temperature Sensing for Use in Leak Detection of Subsea Pipelines

    E-Print Network [OSTI]

    Madabhushi, S.; Elshafie, M. Z. E. B.; Haigh, S. K.

    2014-09-25

    Accurate and rapid detection of leaks is important for subsea oil pipelines to minimize environmental risks and operational/repair costs. Temperature-sensing optical fiber cables can provide economic, near real-time sensing of leaks in subsea oil...

  19. 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

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

    E-Print Network [OSTI]

    Ciocan-Fontanine, Ionut

    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

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-07-11

    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.

  2. IccTA: Detecting Inter-Component Privacy Leaks in Android Apps

    E-Print Network [OSTI]

    McDaniel, Patrick Drew

    IccTA: Detecting Inter-Component Privacy Leaks in Android Apps Li Li, Alexandre Bartel, Tegawend if those are in different components. Since Android applications may leak private data carelessly or maliciously, we propose IccTA, a static taint analyzer to detect privacy leaks among components in Android

  3. 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)

    2006-07-01

    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)

  4. Cofrentes NPP activities on PSA and severe accident analysis

    SciTech Connect (OSTI)

    Suarez, J.; Borondo, L. [IBERDROLA, Madrid (Spain); Garcia, P.J. [UITESA, Madrid (Spain). Nuclear Dept.

    1996-07-01

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power, internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; prioritization of motor operated valves related to GL-89/10, complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; reliability-centered maintenance; maintenance rule support; on-line maintenance support, off-line risk-monitor development, PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of mis-oriented fuel bundle event, etc. About severe accident management, CNPP, as part of the Spanish-BWROG, is currently analyzing the generic products of the US-BWROG AMG in order to generate their specific ones. Also, in this group BWR, the development of tools to simulate accident scenarios beyond core damage will be studied and a training process oriented to warrant the optimum use of new EOP/AMG in accident scenarios will be implemented.

  5. Management of Leaks in Hydrogen Production, Delivery, and Storage Systems

    SciTech Connect (OSTI)

    Rawls, G

    2006-04-27

    A systematic approach to manage hydrogen leakage from components is presented. Methods to evaluate the quantity of hydrogen leakage and permeation from a system are provided by calculation and testing sensitivities. The following technology components of a leak management program are described: (1) Methods to evaluate hydrogen gas loss through leaks; (2) Methods to calculate opening areas of crack like defects; (3) Permeation of hydrogen through metallic piping; (4) Code requirements for acceptable flammability limits; (5) Methods to detect flammable gas; (6) Requirements for adequate ventilation in the vicinity of the hydrogen system; (7) Methods to calculate dilution air requirements for flammable gas mixtures; and (8) Concepts for reduced leakage component selection and permeation barriers.

  6. Accident Tolerant Fuel Analysis

    SciTech Connect (OSTI)

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

    2014-09-01

    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.

  7. Accident tolerant fuel analysis

    SciTech Connect (OSTI)

    Smith, Curtis; Chichester, Heather; Johns, Jesse; Teague, Melissa; Tonks, Michael Idaho National Laboratory; Youngblood, Robert

    2014-09-01

    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.

  8. 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)

    2014-11-15

    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.

  9. Worried about leaks Don't paint before hydrotesting

    SciTech Connect (OSTI)

    Batey, J.E. )

    1993-09-01

    Occasionally, painting before hydrostatic pressure testing is required in petrochemical and other industrial plants. Because some process fluids may be solvents to paint, in-service leakage could occur if the paint masks leakage during hydrotesting. To eliminate unplanned releases, it is important to know whether painting before hydrotesting could really mask leaks at the test pressures typically used in hydrotesting. Unfortunately, very little guidance is provided by national standards or codes, and empirical data are not readily available to support an answer. ASTME 1003-84, Standard Method for Hydrostatic Leak Testing, states that new systems should be tested prior to painting, where practical. However, Sections 1 and 8 of the ASME Boiler and Pressure Vessel Code and B31.1 and B31.3 of the ASME Code for Pressure Piping are silent on this issue. To help resolve this issue, tests were done to determine the effect of paint on leak-tightness during hydrotesting. Pipe samples with through-wall pinholes were fabricated, painted, and then hydrotested.

  10. Cesium-137 deposition and contamination of Japanese soils due to the Fukushima nuclear accident

    E-Print Network [OSTI]

    Jacob, Daniel J.

    Cesium-137 deposition and contamination of Japanese soils due to the Fukushima nuclear accident contamination due to the emission from the Fukushima Daiichi Nuclear Power Plant (NPP) showed up after a massive and severely damaged the Fukushima Daiichi Nuclear Power Plant (NPP). This event led to emissions

  11. Environment/Health/Safety (EHS): Monthly Accident Statistics

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

    Personal Protective Equipment (PPE) Injury Review & Analysis Worker Safety and Health Program: PUB-3851 Monthly Accident Statistics Latest Accident Statistics Accident...

  12. 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.

    2012-06-25

    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.

  13. 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

    1994-06-01

    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.

  14. Accident management for indian pressurized heavy water reactors

    SciTech Connect (OSTI)

    Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

    2006-07-01

    Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

  15. Analysis of accidents during flashing operations 

    E-Print Network [OSTI]

    Obermeyer, Michael Edward

    1993-01-01

    . In this thesis, the relative impacts of flashing signal operation versus regular signal operation were evaluated in several cities and towns in the State of Texas. Analysis were conducted to determine whether an increase in accidents and accident severity...

  16. Specific topics in severe accident management

    SciTech Connect (OSTI)

    Meyer, J.F.; Chung, D.T.; Panciera, V.W.; Traver, L.E.; Humphries, D.S. (SCIENTECH, Inc., Rockville, MD (USA))

    1991-05-01

    This report examines five topical areas of concern to severe accident management. These areas are as follows: Human Factors, Accident Management During Shutdown, Information Needs, Long-term Implications, and Uncertainties. The objective of this report is to assist the NRC in performing its research function and to provide guidance to the industry on accident management strategies, as well as to accident management programs in general. 47 refs., 4 figs., 5 tabs.

  17. A subsea pipeline comprising secondary containment and leak detection

    SciTech Connect (OSTI)

    Kaempen, C.E.

    1996-09-01

    This paper introduces a corrosion-resistant double-wall composite subsea pipe that provides the pipe with secondary containment and leak detection capability. Tables are presented that describe the pressures attainable with the mechanically coupled double-wall composite subsea pipe illustrated in several figures. A description is provided of the construction of the composite subsea pipe and the mechanical coupling assembly used to rapidly connect it during ocean deployment. The paper concludes with a series of questions and answers that provide cost and production information useful for feasibility studies that evaluate factors relating to the replacement of steel subsea pipe with one that promises improved performance.

  18. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QAsource History View NewTexas:Montezuma, Arizona: Energy ResourcesProjectMississippi: EnergyLawrieEdgeLeake County,

  19. UNIVERSITY OF TRENTO ACCIDENT INSURANCE

    E-Print Network [OSTI]

    or freezing; · electrocution; · sunstrokes, hot or cold strokes; · lesions caused by muscular efforts. Sanitary transportation expenses (always included) The Company refunds, up to a liability limit of 500 for the transport from the place of the accident to an equipped Healthcare Institute, the transport among Health

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

    SciTech Connect (OSTI)

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

    2014-07-22

    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.

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2014-05-15

    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.

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-07-30

    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-U Farm

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-12-02

    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.

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-09-10

    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.

  5. 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

  6. Controlling Mole Damage 

    E-Print Network [OSTI]

    Texas Wildlife Services

    2007-03-13

    of underground burrows,comingtothesurfaceonlyrarely,and then often by accident. Because of its secluded lifeunderground,themolehasonlyafewnatur- al enemies. Coyotes, dogs, badgers and skunks dig out a few of them, and occasionally a cat, hawk or owl surprises one...

  7. Method of locating a leaking fuel element in a fast breeder power reactor

    DOE Patents [OSTI]

    Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  8. Bayesian hierarchical models for soil CO{sub 2} flux and leak...

    Office of Scientific and Technical Information (OSTI)

    Bayesian hierarchical models for soil COsub 2 flux and leak detection at geologic sequestration sites Citation Details In-Document Search Title: Bayesian hierarchical models for...

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2013-11-19

    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.

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

    DOE Patents [OSTI]

    Myneni, Ganapati Rao (Yorktown, VA)

    1997-01-01

    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.

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

    DOE Patents [OSTI]

    Myneni, G.R.

    1997-12-30

    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.

  12. You won`t find these leaks with a blower door: The latest in {open_quotes}leaking electricity{close_quotes} in homes

    SciTech Connect (OSTI)

    Rainer, L.; Greenberg, S.; Meier, A.

    1996-08-01

    Leaking electricity is the energy consumed by appliances when they are switched off or not performing their principal functions. Field measurements in Florida, California, and Japan show that leaking electricity represents 50 to 100 Watts in typical homes, corresponding to about 5 GW of total electricity demand in the United States. There are three strategies to reduce leaking electricity: eliminate leakage entirely, eliminate constant leakage and replace with intermittent charge plus storage, and improve efficiency of conversion. These options are constrained by the low value of energy savings-less than $5 per saved Watt. Some technical and lifestyle solutions are proposed. 13 refs., 1 fig., 2 tabs.

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

    SciTech Connect (OSTI)

    Zacharias, T.; Novotny, C.; Bielor, E.

    2012-07-01

    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)

  14. 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.

    1998-08-01

    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).

  15. 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.

    2012-11-11

    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.

  16. System and damage identification of civil structures

    E-Print Network [OSTI]

    Moaveni, Babak

    2007-01-01

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

  17. Accident analysis and DOE criteria

    SciTech Connect (OSTI)

    Graf, J.M.; Elder, J.C.

    1982-01-01

    In analyzing the radiological consequences of major accidents at DOE facilities one finds that many facilities fall so far below the limits of DOE Order 6430 that compliance is easily demonstrated by simple analysis. For those cases where the amount of radioactive material and the dispersive energy available are enough for accident consequences to approach the limits, the models and assumptions used become critical. In some cases the models themselves are the difference between meeting the criteria or not meeting them. Further, in one case, we found that not only did the selection of models determine compliance but the selection of applicable criteria from different chapters of Order 6430 also made the difference. DOE has recognized the problem of different criteria in different chapters applying to one facility, and has proceeded to make changes for the sake of consistency. We have proposed to outline the specific steps needed in an accident analysis and suggest appropriate models, parameters, and assumptions. As a result we feed DOE siting and design criteria will be more fairly and consistently applied.

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

    E-Print Network [OSTI]

    Wheatley, Spencer; Sornette, Didier

    2015-01-01

    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. Mobile Sensor Networks for Leak and Backflow Detection in Water Distribution Systems

    E-Print Network [OSTI]

    Shihada, Basem

    Mobile Sensor Networks for Leak and Backflow Detection in Water Distribution Systems M. Agumbe detection are essential aspects of Water Distribution System (WDS) monitoring. Most existing solutions for leak detection in water distribution systems focus on the placement of expensive static sensors located

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

    SciTech Connect (OSTI)

    Oostrom, Martinus; Wietsma, Thomas W.

    2014-09-30

    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. Internal dissipation and heat leaks in quantum thermodynamic cycles

    E-Print Network [OSTI]

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

    2015-07-06

    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.

  3. Internal dissipation and heat leaks in quantum thermodynamic cycles

    E-Print Network [OSTI]

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

    2015-09-11

    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.

  4. Identifying structural damage from images

    E-Print Network [OSTI]

    Chen, ZhiQiang

    2009-01-01

    feature extraction methods ( A). . . . . . . . . . . .damage feature extraction methods (B). . . . . . . . . . . .a damage feature extraction method and a multi-level

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

    SciTech Connect (OSTI)

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

    2012-03-18

    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.

  6. 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

    2013-01-01

    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...

  7. 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

    2013-06-05

    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.

  8. Farm Fuel Safety Accidents in the handling, use and storage of gasoline, gasohol, diesel fuel, LP-gas and

    E-Print Network [OSTI]

    Tullos, Desiree

    112 Farm Fuel Safety Accidents in the handling, use and storage of gasoline, gasohol, diesel fuel and by keeping fuel storage facilities in top condition. Flammable Liquids and Gases Gasoline, diesel fuel, LP, deterioration or damage. Never store fuel in food or drink containers. When transferring farm fuels, bond

  9. The Accident at Fukushima: What Happened?

    SciTech Connect (OSTI)

    Fujie, Takao

    2012-07-01

    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.

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

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

    for improvement of our integrated safety management system. The techniques and tools utilized in the investigation of "accidents" can be valuable in looking at leading...

  11. 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

    2008-01-01

    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...

  12. Analytical and experimental studies of leak location and environment characterization for the international space station

    SciTech Connect (OSTI)

    Woronowicz, Michael; Blackmon, Rebecca; Brown, Martin; Abel, Joshua; Hawk, Doug; Autrey, David; Glenn, Jodie; Bond, Tim; Buffington, Jesse; Cheng, Edward; Ma, Jonathan; Rossetti, Dino; DeLatte, Danielle; Garcia, Kelvin; Mohammed, Jelila; Montt de Garcia, Kristina; Perry, Radford; Tull, Kimathi; Warren, Eric

    2014-12-09

    The International Space Station program is developing a robotically-operated leak locator tool to be used externally. The tool would consist of a Residual Gas Analyzer for partial pressure measurements and a full range pressure gauge for total pressure measurements. The primary application is to demonstrate the ability to detect NH{sub 3} coolant leaks in the ISS thermal control system. An analytical model of leak plume physics is presented that can account for effusive flow as well as plumes produced by sonic orifices and thruster operations. This model is used along with knowledge of typical RGA and full range gauge performance to analyze the expected instrument sensitivity to ISS leaks of various sizes and relative locations (“directionality”). The paper also presents experimental results of leak simulation testing in a large thermal vacuum chamber at NASA Goddard Space Flight Center. This test characterized instrument sensitivity as a function of leak rates ranging from 1 lb{sub m/}/yr. to about 1 lb{sub m}/day. This data may represent the first measurements collected by an RGA or ion gauge system monitoring off-axis point sources as a function of location and orientation. Test results are compared to the analytical model and used to propose strategies for on-orbit leak location and environment characterization using the proposed instrument while taking into account local ISS conditions and the effects of ram/wake flows and structural shadowing within low Earth orbit.

  13. A comparison of portable and permanent landfill liner leak detection systems

    SciTech Connect (OSTI)

    Taylor, S.B.; White, C.C.; Barker, R.D.

    1999-07-01

    Monitoring of the integrity of electrically non-conductive geomembrane liners installed at waste sites using electrical geophysical techniques has been carried out for a number of years using above-liner leak location surveys and, more recently, below-liner monitoring systems. The authors compare the theoretical response of both types of survey to a hole in a liner and then compare with measurements made in the field. The theoretical leak response indicates that above-liner surveys are sensitive to leaks over a greater area, though both responses result in comparable leak detectability. However, field data suggest that in practice, measurements made on a sparse grid below the liner have the greater sensitivity to certain leaks. This may be due to the differing leak geometries and background conditions present above and below the liner. The results indicate that a sparse below-liner monitoring grid, with its long-term monitoring capabilities, combined with above-liner surveys to pinpoint leaks accurately offer a successful approach to ensuring liner integrity throughout the lifetime of a lined waste site.

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

    SciTech Connect (OSTI)

    Gauntt, Randall O.

    2010-04-01

    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.

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2014-01-08

    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.

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

    SciTech Connect (OSTI)

    Girardot, Crystal L.; Harlow, Donald G.

    2014-09-04

    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.

  17. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Hoover, M.D.; Newton, G.J. [Inhalation Toxicology Research Inst., Albuquerque, NM (United States); Farrell, R.F. [Dept. of Energy, Carlsbad, NM (United States)

    1996-06-01

    This qualitative hazard evaluation systematically assessed potential doses to workers during postulated accident conditions at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). Postulated accidents included the spontaneous ignition of a waste drum, puncture of a waste drum by a forklift, dropping of a waste drum from a forklift, and simultaneous dropping of seven drums during a crane failure. The descriptions and estimated frequencies of occurrence for these accidents were developed by the Hazard and Operability Study for CH TRU Waste Handling System (WCAP 14312). The estimated materials at risk, damage ratios, airborne release fractions and respirable fractions for these accidents were taken from the 1995 Safety Analysis Report (SAR) update and from the DOE handbook Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities (DOE-HDBK-3010-94). A Monte Carlo simulation was used to estimate the range of worker exposures that could result from each accident. Guidelines for evaluating the adequacy of defense-in-depth for worker protection at WIPP were adopted from a scheme presented by the International Commission on Radiological Protection in its publication on Protection from Potential Exposure: A Conceptual Framework (ICRP Publication 64). Probabilities of exposures greater than 5, 50, and 300 rem were less than 10{sup -2}, 10{sup -4}, and 10{sup -6} per year, respectively. In conformance with the guidance of DOE standard 3009-94, Appendix A (draft), we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposure under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, as well as members of the public and the environment.

  18. Does Daylight Savings Time Affect Traffic Accidents

    E-Print Network [OSTI]

    Deen, Sophia 1988-

    2012-04-20

    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...

  19. Light-water reactor accident classification

    SciTech Connect (OSTI)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  20. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect (OSTI)

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

    2012-11-01

    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.

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

    Office of Environmental Management (EM)

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

  2. 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...

  3. Type B Accident Investigation Board Report on the October 8,...

    Energy Savers [EERE]

    Bettis Atomic Power Laboratory Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 Type B Accident Investigation Board Report on...

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

    Energy Savers [EERE]

    Type 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...

  5. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the...

  6. Accident Investigation of the February 7, 2013, Scissor Lift...

    Energy Savers [EERE]

    Lift Accident in the West Hackberry Brine Tank-14 Resulting in Injury, Strategic Petroleum Reserve West Hackberry, LA Accident Investigation of the February 7, 2013, Scissor...

  7. 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...

  8. 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...

  9. 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)

    1995-01-01

    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.

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

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    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.

  11. Detecting Leaks and Sensor Biases by Recursive Identification with Forgetting Factors

    E-Print Network [OSTI]

    Marquez, Horacio J.

    . In this paper, a process model is proposed to describe boiler tube leak problem. Based on this model, least bias. The application in a boiler system shows that the proposed methods can detect the boiler tube

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

    SciTech Connect (OSTI)

    Jerry Myers

    2005-04-15

    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.

  13. PLC Software Program for Leak Detector Station A1 SALW-LD-ST-A1

    SciTech Connect (OSTI)

    KOCH, M.R.

    2001-01-25

    This document describes the software program for the programmable logic controller for the leak detector station ''SALW-LD-ST-A1''. The appendices contains a copy of the printout of the software program.

  14. 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

    2013-06-01

    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.

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

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

    And Detection of Leaks (IntelliCLAD) An exaggerated representation of the IntelliCLAD coating in action. A breach in the coating produces a color change at the point of failure...

  16. Safety evaluation of MHTGR licensing basis accident scenarios

    SciTech Connect (OSTI)

    Kroeger, P.G.

    1989-04-01

    The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential accident scenarios that might lead to fuel failures due to excessive core temperatures and/or to vessel damage, due to excessive vessel temperatures. The design basis accident scenario leading to the highest vessel temperatures is the depressurized core heatup scenario without any forced cooling and with decay heat rejection to the passive Reactor Cavity Cooling System (RCCS). This scenario was evaluated, including numerous parametric variations of input parameters, like material properties and decay heat. It was found that significant safety margins exist, but that high confidence levels in the core effective thermal conductivity, the reactor vessel and RCCS thermal emissivities and the decay heat function are required to maintain this safety margin. Severe accident extensions of this depressurized core heatup scenario included the cases of complete RCCS failure, cases of massive air ingress, core heatup without scram and cases of degraded RCCS performance due to absorbing gases in the reactor cavity. Except for no-scram scenarios extending beyond 100 hr, the fuel never reached the limiting temperature of 1600/degree/C, below which measurable fuel failures are not expected. In some of the scenarios, excessive vessel and concrete temperatures could lead to investment losses but are not expected to lead to any source term beyond that from the circulating inventory. 19 refs., 56 figs., 11 tabs.

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

    SciTech Connect (OSTI)

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

    2013-05-24

    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.

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

    E-Print Network [OSTI]

    Ramaratham, Srivatsan

    2008-10-10

    of leaks. Brake systems in trucks are crucial for ensuring the safety of vehicles and passengers on the roadways. Most trucks in the US are equipped with S-cam drum brake systems and they are sensitive to maintenance. Brake defects such as leaks are a major... and schematic of operation. . . . . . . . . . . . . 7 3 A typical drum brake assembly. . . . . . . . . . . . . . . . . . . . . . 8 4 Front and rear brake chambers. . . . . . . . . . . . . . . . . . . . . . 8 5 Automatic slack adjuster construction...

  19. if it is a gas leak, do not activate building alarms, use mobile phones, hand held radios, electronic equipment or light flammable material!

    E-Print Network [OSTI]

    Hickman, Mark

    gas leak gas leak if it is a gas leak, do not activate building alarms, use mobile phones, hand held radios, electronic equipment or light flammable material! 1. If you discover a Gas Leak, shout and check that the nearest gas isolator switch is off. 4. Evacuate the building immediately, avoiding

  20. Covered Biodegradable Stent: New Therapeutic Option for the Management of Esophageal Perforation or Anastomotic Leak

    SciTech Connect (OSTI)

    Cerna, Marie; Koecher, Martin Valek, Vlastimil; Aujesky, Rene; Neoral, Cestmir; Andrasina, Tomas; Panek, Jiri; Mahathmakanthi, Shankari

    2011-12-15

    Purpose: This study was designed to evaluate our experience with the treatment of postoperative anastomotic leaks and benign esophageal perforations with covered biodegradable stents. Materials and Methods: From 2008 to 2010, we treated five men with either an anastomotic leak or benign esophageal perforation by implanting of covered biodegradable Ella-BD stents. The average age of the patients was 60 (range, 38-74) years. Postoperative anastomotic leaks were treated in four patients (1 after esophagectomy, 1 after resection of diverticulum, 2 after gastrectomy). In one patient, perforation occurred as a complication of the treatment of an esophageal rupture (which occurred during a balloon dilatation of benign stenosis) with a metallic stent. Results: Seven covered biodegradable stents were implanted in five patients. Primary technical success was 100%. Clinical success (leak sealing) was achieved in four of the five patients (80%). Stent migration occurred in three patients. In two of these patients, the leak had been sealed by the time of stent migration, therefore no reintervention was necessary. In one patient an additional stent had to be implanted. Conclusion: The use of biodegradable covered stents for the treatment of anastomotic leaks or esophageal perforations is technically feasible and safe. The initial results are promising; however, larger number of patients will be required to evaluate the capability of these biodegradable stents in the future. The use of biodegradable material for coverage of the stent is essential.

  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)

    1994-09-30

    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. RADIATION DAMAGE OF GERMANIUM DETECTORS

    E-Print Network [OSTI]

    Pehl, Richard H.

    2011-01-01

    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

  3. Controlling Beaver Damage 

    E-Print Network [OSTI]

    Texas Wildlife Services

    2007-03-13

    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....

  4. Controlling Feral Hog Damage 

    E-Print Network [OSTI]

    Texas Wildlife Services

    2008-04-15

    This publication discusses the distribution of feral hogs as well as their habitats, food habits and reproduction. Feral hogs can damage crops and kill lambs and kid goats. Methods of control are also explained....

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

    SciTech Connect (OSTI)

    David Cist; Alan Schutz

    2005-03-30

    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.

  6. Leak Detection and H2 Sensor Development for Hydrogen Applications

    SciTech Connect (OSTI)

    Brosha, Eric L.

    2012-07-10

    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.

  7. 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

  8. Causal Reasoning About Aircraft Accidents Peter B. Ladkin

    E-Print Network [OSTI]

    Ladkin, Peter B.

    , rigorous causal reasoning in the analysis of air transportation accidents can improve our understanding

  9. 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

  10. Severe fuel-damage scoping test performance. [PWR

    SciTech Connect (OSTI)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle.

  11. Release fractions for Rocky Flats specific accidents

    SciTech Connect (OSTI)

    Weiss, R.C.

    1992-09-01

    As Rocky Flats and other DOE facilities begin the transition process towards decommissioning, the nature of the scenarios to be studied in safety analysis will change. Whereas the previous emphasis in safety accidents related to production, now the emphasis is shifting to accidents related tc decommissioning and waste management. Accident scenarios of concern at Rocky Flats now include situations of a different nature and different scale than are represented by most of the existing experimental accident data. This presentation will discuss approaches@to use for applying the existing body of release fraction data to this new emphasis. Mention will also be made of ongoing efforts to produce new data and improve the understanding of physical mechanisms involved.

  12. Crediting Tritium Deposition in Accident Analysis

    SciTech Connect (OSTI)

    Murphy, C.E. Jr.

    2001-06-20

    This paper describes the major aspects of tritium dispersion phenomenology, summarizes deposition attributes of the computer models used in the DOE Complex for tritium dispersion, and recommends an approach to account for deposition in accident analysis.

  13. A systems approach to food accident analysis

    E-Print Network [OSTI]

    Helferich, John D

    2011-01-01

    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. ...

  14. Structural assessment of accident loads

    SciTech Connect (OSTI)

    Wagenblast, G.R., Westinghouse Hanford

    1996-05-28

    Structural assessments were made for specific accident loads for specific catch, receiver, and storage tanks. The evaluation herein represents level-of-effort order-of-magnitude estimates of limiting loads that would lead to collapse or rupture of the tank and unmitigated loss of confinement for the waste. Structural capacities were established using failure criteria. Compliance with codes such as ACI, ASCE, ASME, RCRA, UBC, WAC, and DOE Orders was `NOT` maintained. Normal code practice is to prevent failure with margins consistent with expected variations in loads and strengths and confidence in analysis techniques. The evaluation herein represent estimates of code limits without code load factors or code strength reduction factors, and loading beyond such a limit is considered as an onset of some failure mode. The exact nature of the failure mode and its relation to a safe condition is a judgment of the analyst. Consequently, these `RESULTS SHALL NOT BE USED TO ESTABLISH OPERATING OR SAFETY LOAD LIMITS FOR THESE TANKS`.

  15. Million Cu. Feet Percent of National Total

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    as known volumes of natural gas that were the result of leaks, damage, accidents, migration, andor blow down. Notes: Totals may not add due to independent rounding. Prices are...

  16. Uncertainties and severe-accident management

    SciTech Connect (OSTI)

    Kastenberg, W.E. (Univ. of California, Los Angeles (United States))

    1991-01-01

    Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies.

  17. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    SciTech Connect (OSTI)

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  18. Evaluation of EBR-II driver-fuel elements following an unprotected station blackout accident

    SciTech Connect (OSTI)

    Chang, L.K.; Bottcher, J.H.

    1986-01-01

    One of the current design objectives for a liquid metal reactor (LMR) is the inherent shutdown-cooling capability of the reactor, such that the reactor itself can safely reduce power following a total loss of pump power without activating the reactor shutdown system (RSS). Following a loss-of-flow (LOF) accident and a failure of RSS, in EBR-II, reactor core damage and plant restartability is of considerable interest. In the LOF event, high temperature in the reactor causes negative reactivity feedback that reduces reactor power. After an accident, reactor fuel performance is one of the factors used to assess the restartability of the plant. A thermal-hydraulic-neutronic analysis was performed to determine the response of the plant and the temperature of individual subassemblies. These temperatures were then used to assess the damage to driver fuel elements caused by the station blackout accident. The maximum depth of cladding wastage from molten eutectic at temperatures >715/sup 0/C was found to be 0.0053 mm for the hottest subassembly; this value is considerably less than the 0.28 mm cladding thickness. 12 refs.

  19. Role of Passive Safety Systems in Severe Accidents Prevention for Advanced WWER-1000 Reactor Plants

    SciTech Connect (OSTI)

    Bukin, N.V.; Fil, N.S.; Shumsky, A.M. [EDO 'Gidropress', 21 Ordzhonikidze str., Podolsk, Moscow Region, RU-142103 (Russian Federation)

    2004-07-01

    Role of new safety systems applied in advanced WWER-1000 (passive residual heat removal system, SPOT and passive core flooding system, HA-2) in severe accident prevention is considered in the paper. The following typical beyond-design accidents (BDBAs) that essentially determine the design basis of the above passive systems are considered in the paper: - station blackout; - LB LOCA (double-ended cold leg break 850 mm diameter) with station blackout. The domestic DINAMIKA-97 and TETCH-M-97 codes developed by EDO 'Gidropress' were used for the analyses. Besides, some supporting calculations have been performed by new Russian KORSAR code and western RELAP5/MOD3.2 and ATHLET 1.2A codes. The analysis of station blackout accident without operation of new passive systems have shown the exceeding of the maximum design limit of fuel rod damage already in 2-2,5 h after initiating event. Operation of SPOT system prevents any core damage during the BDBA under consideration. The analysis have also demonstrated that operation of new passive safety systems (SPOT and HA-2) ensures the effective core cooling within required period of time. This ensures essentially decreased probability of severe core degradation. (authors)

  20. 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

    1997-04-01

    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.

  1. 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

    2011-07-27

    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.

  2. Correlating cookoff violence with pre-ignition damage.

    SciTech Connect (OSTI)

    Wente, William Baker; Hobbs, Michael L.; Kaneshige, Michael Jiro

    2010-03-01

    Predicting the response of energetic materials during accidents, such as fire, is important for high consequence safety analysis. We hypothesize that responses of ener-getic materials before and after ignition depend on factors that cause thermal and chemi-cal damage. We have previously correlated violence from PETN to the extent of decom-position at ignition, determined as the time when the maximum Damkoehler number ex-ceeds a threshold value. We seek to understand if our method of violence correlation ap-plies universally to other explosive starting with RDX.

  3. Standard specification for leak detector solutions intended for use on brasses and other copper alloys

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This specification covers the requirements for leak detector solutions suitable for use in checking the leakage of valves, pipes, fittings, joints, and so forth of a pressurized gas system fabricated from brasses and other copper alloys. 1.2 This specification deals with the stress corrosion cracking aspect of leak detector solutions. The effectiveness, chemical, physical and mechanical properties of leak detector solutions are not within the scope of this specification. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and to determine the applicability of regulatory limitations prior to use.

  4. Saltwell Leak Detector Station Programmable Logic Controller (PLC) Software Configuration Management Plan (SCMP)

    SciTech Connect (OSTI)

    WHITE, K.A.

    2000-11-28

    This document provides the procedures and guidelines necessary for computer software configuration management activities during the operation and maintenance phases of the Saltwell Leak Detector Stations as required by HNF-PRO-309, Rev. 1, Computer Software Quality Assurance, Section 2.4, Software Configuration Management. The software configuration management plan (SCMP) integrates technical and administrative controls to establish and maintain technical consistency among requirements, physical configuration, and documentation for the Saltwell Leak Detector Station Programmable Logic Controller (PLC) software during the Hanford application, operations and maintenance. This SCMP establishes the Saltwell Leak Detector Station PLC Software Baseline, status changes to that baseline, and ensures that software meets design and operational requirements and is tested in accordance with their design basis.

  5. Interfacing systems loss-of-coolant accident in Oconee-1 pressurized water reactor

    SciTech Connect (OSTI)

    Nassersharif, B.

    1984-01-01

    The primary system of a pressurized water reactor (PWR) operates at a relatively high pressure (15.5 MPa, 2250 psia) and consists of piping and components designed to withstand these pressures. The low-pressure-injection system (LPIS) connects to the primary system but possesses low-pressure piping passing outside the containment. Therefore, a potential exists for a loss-of-coolant accident (LOCA) outside the containment and concurrent damage to systems needed to cope with this problem. The emergency core-cooling system (ECCS) is assumed to be available for this event. A set of calculations were performed using the TRAC-PF1 code and a model of the Oconee-1 PWR to investigate the consequences of, and possible operator actions for, such an accident scenario.

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

    SciTech Connect (OSTI)

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

    2007-11-01

    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.

  7. 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

    2005-05-01

    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.

  8. 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.

    1986-10-01

    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.

  9. Analysis of Severe Accident Management Strategy for a BWR-4 Nuclear Power Plant

    SciTech Connect (OSTI)

    Wang, T.-C.; Wang, S.-J.; Teng, J.-T

    2005-12-15

    The Chinshan nuclear power plant (NPP) is a Mark-I boiling water reactor (BWR) NPP located in northern Taiwan. The Chinshan NPP severe accident management guidelines (SAMGs) were developed based on the BWR Owners Group Emergency Procedure Guidelines/Severe Accident Guidelines and were developed at the end of 2003. The MAAP4 code has been used as a tool to validate the SAMG strategies. The development process and characteristics of the Chinshan SAMGs are described. The T{sub 5}U{sub t}X{sub C} sequence, the highest core damage frequency in the probabilistic risk assessment insight of the Chinshan NPP, is cited as a reference case for SAMG validation. Not all safety injection systems are operated in the T{sub 5}U{sub t}X{sub C} sequence. The severe accident progression is simulated, and the entry condition of the SAMGs is described. Then, the T{sub 5}U{sub t}X{sub C} sequence is simulated with reactor pressure vessel (RPV) depressurization. Mitigation actions based on the Chinshan NPP SAMGs are then applied to demonstrate the effectiveness of the SAMGs. Sensitivity studies on RPV depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. Based on MAAP4 calculation and the default values of the parameters calculating the severe accident phenomena, the result shows that RPV depressurization before the reactor water level reaches one-fourth of the core water level can prevent the core from damage in the T{sub 5}U{sub t}X{sub C} sequence. The flow rate of two control rod drive pumps is enough to maintain the reactor water level above the top of active fuel and cool down the core in the T{sub 5}U{sub t}X{sub C} sequence without operator action.

  10. 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

    2000-05-01

    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.

  11. 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

    2013-09-01

    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.

  12. 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

  13. 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

    2014-02-01

    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).

  14. RADIATION DAMAGE OF GERMANIUM DETECTORS

    E-Print Network [OSTI]

    Pehl, Richard H.

    2011-01-01

    occurring within the detector, radiation may also change theLBL-7967 RADIATION DAMAGE OF GERMANIUM DETECTORS Richard H.LBL-7967 RADIATION DAMAGE OF GERMANIUM DETECTORS* Richard H.

  15. Development of the severe accident management guidelines (SAMG) for Ulchin Nuclear Power Plant Unit 3, 4, 5 and 6

    SciTech Connect (OSTI)

    Kim, Hyeong T.; Yoo, Hojong; Lim, Hyuk Soon; Park, Jong W.; Lim, Woosang; Oh, Seung Jong [Korea Hydro and Nuclear Power Co., Ltd., 103-16 Munji-Dong, Yusung-Gu, Daejeon, 305-380 (Korea, Republic of); Chung, Chang Hyun [Seoul National University (Korea, Republic of); Lee, Byung Chul [Future and Challenges, Inc (Korea, Republic of)

    2004-07-01

    This paper describes the development process of the severe accident management guidelines (SAMG) for Units 3, 4, 5 and 6 of Ulchin Nuclear Power Plant. The units are Korean Standard Nuclear Power (KSNP) plant, 1000 MWe class pressurized water reactor (PWR) with two loops of primary coolant system. The severe accident management guidelines for the units have been completed in 2002. The generic severe accident management guidance for Korean Standard Nuclear Power Plant has been used as the basis when developing Ulchin severe accident management guideline. Result of probabilistic safety assessment (PSA) for each unit was reviewed to integrate its insight into the SAMG. It indicates that each unit has a balanced design to any specific initiating events for core damage. Seven severe accident management strategies are applied in Ulchin SAMG. Seven strategies are (1) Inject into the steam generator (2) De-pressurize the RCS (3) Inject into the RCS (4) Inject into the containment (5) Control the fission product release into environment (6) Control the containment pressure and temperature and (7) Control hydrogen concentration in the containment. The range and capability of essential instrument for performing the strategies are assessed. Computational aids are developed to complement the unavailable instrument during the accident and to assist the operator's decision choosing strategies. To examine the ability of the SAMG to fulfill its intended function, small loss of coolant accident (SLOCA) with the failure of safety injection was selected as a reference scenario. The scenario was analyzed using MAAP code. The evaluation of the SAMG using this sequence has been successfully completed. (authors)

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

    E-Print Network [OSTI]

    McKemie, Martha Susan

    1979-01-01

    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...

  17. VBGVBG3/7/2006MODIFIED LEAK DETECTOR BRACKET1 2. WELDING & INSPECTIONS SHALL BE PERFORMED

    E-Print Network [OSTI]

    McDonald, Kirk

    VBGVBG3/7/2006MODIFIED LEAK DETECTOR BRACKET1 APPROVED 2. WELDING & INSPECTIONS SHALL BE PERFORMED 4. MATERIAL CERTIFICATIONS REQUIRED 3. ALL WELDS SHALL BE DYE PENETRANT INSPECTED IN ACCORDANCE WITH AWS D.1.6 1 1. ALL WELDS MUST BE WATERTIGHT NOTES BYDATEDESCRIPTIONREV ITEM NO. QTY. DESCRIPTION

  18. 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

  19. 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

  20. 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

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

    E-Print Network [OSTI]

    Fernandez, Eduardo

    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

  2. Method for sealing remote leaks in an enclosure using an aerosol

    DOE Patents [OSTI]

    Modera, Mark P. (Piedmont, CA); Carrie, Francois R. (Lyons, FR)

    1999-01-01

    The invention is a method and device for sealing leaks remotely by means of injecting a previously prepared aerosol into the enclosure being sealed according to a particular sealing efficiency defined by the product of a penetration efficiency and a particle deposition efficiency. By using different limits in the relationship between penetration efficiency and flowrate, the same method according the invention can be used for coating the inside of an enclosure. Specifically the invention is a method and device for preparing, transporting, and depositing a solid phase aerosol to the interior surface of the enclosure relating particle size, particle carrier flow rate, and pressure differential, so that particles deposited there can bridge and substantially seal each leak, with out providing a substantial coating at inside surfaces of the enclosure other than the leak. The particle size and flow parameters can be adjusted to coat the interior of the enclosure (duct) without substantial plugging of the leaks depending on how the particle size and flowrate relationships are chosen.

  3. Automatic Detection of Inter-application Permission Leaks in Android Applications

    E-Print Network [OSTI]

    Wallach, Dan

    Automatic Detection of Inter-application Permission Leaks in Android Applications Drago¸s Sb Department of Computer Science, Rice University 2 IBM Watson Research Center Abstract The Android operating, called the Intent mechanism. In this paper we develop techniques for statically detecting Android

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

    E-Print Network [OSTI]

    Beckermann, Christoph

    used by foundries to detect solidification shrinkage defects in steel castings is the Niyama criterion to Predict Shrinkage-Related Leaks in High-Nickel Steel and Nickel-Based Alloy Castings," in Proceedings shrinkage that is not visible on a standard radiographic film), other casting features, or some combination

  5. 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.

    2012-12-20

    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).

  6. Tracer Gas as a Practical Field Diagnostic Tool for Assessing Duct System Leaks 

    E-Print Network [OSTI]

    Cummings, J. B.

    1989-01-01

    diagnostic tools for detecting and locating leaks in the air distribution system. The tracer gas tests described are a good complement to these tools in the detection, location, and measurement of duct leakage. Testing for house infiltration once with the air...

  7. The BWR lower head response during a large-break LOCA with core damage

    SciTech Connect (OSTI)

    Alammar, M.A. [GPU Nuclear Corp., Parsippany, NJ (United States)

    1996-12-31

    Some of the important issues in severe accident management guidelines development deal with estimating the time to lower head vessel failure after core damage and the time window available for water injection that would prevent vessel failure. These issues are obviously scenario dependent, but bounding estimates are needed. The scenario chosen for this purpose was a design-basis accident (DBA) loss-of-coolant accident (LOCA) because it was one of the contributors to the Oyster Creek containment failure frequency. Oyster Creek is a 1930-MW(thermal) boiling water reactor (BWR)-2. The lower head response models have improved since the Three Mile Island unit 2 (TMI-2) vessel investigation project (VIP) results became known, specifically the addition of rapid- and slow-cooling models. These mechanisms were found to have taken place in the TMI-2 lower head during debris cooldown and were important contributors in preventing vessel failure.

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

    Energy Savers [EERE]

    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...

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

    SciTech Connect (OSTI)

    Hill, Robin L.; Conrady, Matthew M.

    2010-06-30

    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.

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

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M.

    2014-03-01

    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.

  11. 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.

    2014-03-01

    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

  12. 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)

    1996-02-01

    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.

  13. 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

  14. 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

  15. Severe Accident Test Station Activity Report

    SciTech Connect (OSTI)

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  16. Characterization of a nuclear accident dosimeter 

    E-Print Network [OSTI]

    Burrows, Ronald Allen

    1995-01-01

    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...

  17. Accident or Design Taeil A. Bai

    E-Print Network [OSTI]

    Bai, Taeil

    to this principle, a life-giving factor lies at the center of the whole machinery and design of the world. Here weAccident or Design Taeil A. Bai Stanford University, Stanford, CA 94305 In a companion article find the word "design," which has been expelled from biological scholarship by the biologists. Here we

  18. 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

  19. Louisiana Forest Products Lab 1 Accidents in the Primary &

    E-Print Network [OSTI]

    Louisiana Forest Products Lab 1 Accidents in the Primary & Secondary Forest Products Industry Center #12;Louisiana Forest Products Lab 2 Abitibi Paper Co. Camp 40 Thunder Bay, Ontario #12;Louisiana Forest Products Lab 3 Accidents in Forest Products Industry Accident Statistics Primary industry

  20. Model based detection of hydrogen leaks in a fuel cell stack Ari Ingimundarson and Anna G. Stefanopoulou and Denise McKay

    E-Print Network [OSTI]

    Stefanopoulou, Anna

    will depend on the composition of the gas where the leak takes place. Two approaches are presented here but takes into account the natural leak of the stack and humidity. Hydrogen leak detection without using. Hydrogen has the lowest molecular weight and viscosity of any gas. Its properties make it have a faster

  1. Controlling Armadillo Damage 

    E-Print Network [OSTI]

    Texas Wildlife Services

    2007-03-13

    of overlapping rings. The under- parts are covered with soft skin and a few long hairs. The armadillo is about the size of an opos- sum. Its front feet are well adapted for digging. Tracks made by an armadillo appear to have been made by a three-toed animal... and other invertebrates, as well as on small amounts of fruit and vegetable matter such as berries and tender roots. Damage Although armadillos are beneficial because they eat insects and other invertebrates, they sometimes become a nuisance by digging...

  2. Comparisons of the SCDAP computer code with bundle data under severe accident conditions. [PWR; BWR

    SciTech Connect (OSTI)

    Allison, C.M.; Beers, G.H.

    1983-01-01

    The SCDAP computer code, which is being developed under the sponsorship of the United States Nuclear Regulatory Commission, models the progression of light water reactor core damage including core heatup, core disruption and debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data. Comparisons between SCDAP calculations and the experimental data showed good agreement. Calculated and measured bundle temperatures for SFD-ST were within 200 K for the entire bundle and within 20 K for maximum cladding temperatures. For ESSI-2, calculated and measured maximum cladding temperatures were within 50 K, and the extensive liquefaction and relocation that was calculated was in agreement with experimental results.

  3. 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.

    1999-10-01

    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.

  4. 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

    2012-12-12

    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.

  5. 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))

    1990-10-01

    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.

  6. Nowcasting Disaster Damage

    E-Print Network [OSTI]

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

    2015-01-01

    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...

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

    SciTech Connect (OSTI)

    Wu, T

    2008-04-30

    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.

  8. 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

  9. Radiation Damage/Materials Modification

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

    ion irradiation is an important tool for studying radiation damage effects Materials in a nuclear reactor are exposed to extreme temperature and radiation conditions that degrade...

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

    SciTech Connect (OSTI)

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

    2014-12-22

    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.

  11. A modified heat leak test facility employing a closed-cycle helium refrigerator

    SciTech Connect (OSTI)

    Boroski, W.N.

    1996-01-01

    A Heat Leak Test Facility (HLTF) has been in use at Fermilab for many years. The apparatus has successfully measured the thermal performance of a variety of cryostat components under simulated operating conditions. While an effective tool in the cryostat design process, the HLTF has several limitations. Temperatures are normally fixed at cryogen boiling points and run times are limited to cryogen inventory. Moreover, close personnel attention is required to maintain system inventories and sustain system equilibrium. To provide longer measurement periods without perturbation and to minimize personnel interaction, a new heat leak measurement facility (HLTF-2) has been designed that incorporates a closed-cycle helium refrigerator. The two-stage refrigerator provides cooling to the various temperature stations of the HLTF while eliminating the need for cryogens. Eliminating cryogen inventories has resulted in a reduction of the amount of direct personnel attention required.

  12. US Department of Energy Chernobyl accident bibliography

    SciTech Connect (OSTI)

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

    1992-04-01

    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.

  13. Risk Estimation Methodology for Launch Accidents.

    SciTech Connect (OSTI)

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

    2014-02-01

    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.

  14. Method of assessing severe accident management strategies

    SciTech Connect (OSTI)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.; Jae, M.; Lim, H.; Milici, T.; Park, H.; Swider, J.; Xing, L.; Yu, D. (Univ. of California, Los Angeles (United States))

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems, and actions to prevent or mitigate a severe accident. A significant number of probabilistic safety assessments (PSAs) have been completed that yield the principal plant vulnerabilities. These vulnerabilities can be categorized as (1) dominant sequences with respect to core-melt frequency. (2) dominant sequences with respect to various risk measures. (3) dominant threats that challenge safety functions. (4) dominant threats with respect to failure of safety systems. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainties in key phenomena, operator behavior, system availability and behavior, and available information. This paper presents a methodology for assessing severe accident management strategies given the key uncertainties delineated at two workshops held at the University of California, Los Angeles. Based on decision trees and influence diagrams, the methodology is currently being applied to two case studies: cavity flooding in a pressurized water reactor (PWR) to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent vessel and/or containment failure.

  15. Hanford waste tank bump accident analysis

    SciTech Connect (OSTI)

    MALINOVIC, B.

    2003-03-21

    This report provides a new evaluation of the Hanford tank bump accident analysis (HNF-SD-Wh4-SAR-067 2001). The purpose of the new evaluation is to consider new information and to support new recommendations for final safety controls. This evaluation considers historical data, industrial failure modes, plausible accident scenarios, and system responses. A tank bump is a postulated event in which gases, consisting mostly of water vapor, are suddenly emitted from the waste and cause tank headspace pressurization. A tank bump is distinguished from a gas release event in two respects: First, the physical mechanism for release involves vaporization of locally superheated liquid, and second, gases emitted to the head space are not flammable. For this reason, a tank bump is often called a steam bump. In this report, even though non-condensible gases may be considered in bump models, flammability and combustion of emitted gases are not. The analysis scope is safe storage of waste in its current configuration in single-shell tanks (SSTs) and double-shell tanks (DSTs). The analysis considers physical mechanisms for tank bump to formulate criteria for bump potential, application of the criteria to the tanks, and accident analysis of bump scenarios. The result of consequence analysis is the mass of waste released from tanks for specific scenarios where bumps are credible; conversion to health consequences is performed elsewhere using standard Hanford methods (Cowley et al. 2000). The analysis forms a baseline for future extension to consider waste retrieval.

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

    SciTech Connect (OSTI)

    RUCKER DF; FINK JB; LOKE MH; MYERS DA

    2009-11-05

    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.

  17. 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)

    2013-11-06

    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.

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

    SciTech Connect (OSTI)

    Powers, Dana Auburn

    2009-04-01

    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.

  19. Helium bombardment leak testing of the closure disk weld for MC2949, MC3004, and MC3095 pyrotechnic devices

    SciTech Connect (OSTI)

    Dudley, W.A.

    1980-03-31

    A helium bombardment leak test procedure was developed to determine the leak level of the closure disk weld performed on three nearly identical pyrotechnic actuators. The inspection procedure is capable of leak testing any of the three product types at a rate better than 120 units per 8-hr work shift. Testing is performed on a 100% sample plan and employs a go/no-go bombardment leak rate acceptance specification of 3 x 10/sup -9/ atm-cm/sup 3/-sec/sup -1/. In addition to the current test procedure and results, this report includes a description of procedure and results associated with the test as initially performed. Other applications of the current technique are also listed.

  20. 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)

    2012-07-01

    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)

  1. Shock Initiation of Damaged Explosives

    SciTech Connect (OSTI)

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

    2009-10-22

    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.

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

    SciTech Connect (OSTI)

    Daugherty, W.

    2014-08-06

    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.

  3. Developing and assessing accident management plans for nuclear power plants

    SciTech Connect (OSTI)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

    1992-07-01

    This document is the second of a two-volume NUREG/CR that discusses development of accident management plans for nuclear power plants. The first volume (a) describes a four-phase approach for developing criteria that could be used for assessing the adequacy of accident management plans, (b) identifies the general attributes of accident management plans (Phase 1), (c) presents a prototype process for developing and implementing severe accident management plans (Phase 2), and (d) presents criteria that can be used to assess the adequacy of accident management plans. This volume (a) describes results from an evaluation of the capabilities of the prototype process to produce an accident management plan (Phase 3) and (b), based on these results and preliminary criteria included in NUREG/CR-5543, presents modifications to the criteria where appropriate.

  4. 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

    1993-09-01

    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.

  5. Assessment of light water reactor accident management programs and experience

    SciTech Connect (OSTI)

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  6. Use of probabilistic safety analyses in severe accident management

    SciTech Connect (OSTI)

    Neogy, P.; Lehner, J.

    1991-01-01

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs.

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

    Energy Savers [EERE]

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

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

    Energy Savers [EERE]

    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...

    Energy Savers [EERE]

    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 Illnesses...

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

    Energy Savers [EERE]

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

  11. Accident Investigation of the August 21, 2012, Contamination...

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

    August 21, 2012, Contamination Incident at the Los Alamos Neutron Science Center at the Los Alamos National Laboratory Accident Investigation of the August 21, 2012, Contamination...

  12. Type B Accident Investigation of the October 9, 2008 Employee...

    Energy Savers [EERE]

    October 9, 2008 Employee Injured when Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office Type B Accident...

  13. Type B Accident Investigation Board Report for the January 11...

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

    a serious injury to his right hand while operating a table saw. In conducting its investigation, the Accident Investigation Board (the Board) used various analytical techniques,...

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

    Energy Savers [EERE]

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

  15. Summary of a workshop on severe accident management for PWRs

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of strategy, there may be several options available to the operator; and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainty includes operator, system and instrument behavior during severe accidents. During the period May 15--17, 1990 a workshop was held at the University of California, Los Angeles, to address these uncertainties for pressurized water reactors (PWRs). This report contains a summary of the workshop proceedings.

  16. Summary of a workshop on severe accident management for BWRs

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of 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 instrument behavior during an accident. During the period September 26--28, 1990, a workshop was held at the University of California, Los Angeles, to address these uncertainties for Boiling Water Reactors (BWRs). This report contains a summary of the workshop proceedings.

  17. Accident Investigation of the October 1, 2013, Tice Electric...

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

    22, 2013 On October 2, 2013, at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was convened to investigate an...

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

    Energy Savers [EERE]

    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 Fall...

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

    Office of Environmental Management (EM)

    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...

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

    Office of Environmental Management (EM)

    part of a Washington Closure Hanford, LLC (WCH) team of craft personnel preparing a bridge crane for removal from the 336 Building. Type B Accident Investigation At Washington...

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

    Office of Environmental Management (EM)

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

  2. 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...

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

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

    August 17, 2004 On May 8, 2004, at approximately 11:00 am, an exothermic metal reaction (exothermic reaction) accident occurred during heating of surplus activated sodium shields...

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

    E-Print Network [OSTI]

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

    2007-01-01

    pp. 283-291. Estimating Pedestrian Accident Exposure: Draftand J. Thiran. Counting Pedestrians in Video Sequences UsingPartnership (CLP) Automatic Pedestrian Counting Trial. Stage

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

    Office of Environmental Management (EM)

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

  6. Improvement of Design Codes to Account for Accident Thermal Effects...

    Office of Environmental Management (EM)

    IMPROVEMENT OF DESIGN CODES TO ACCOUNT FOR ACCIDENT THERMAL EFFECTS ON SEISMIC PERFORMANCE Amit H. Varma, Kadir Sener, Saahas Bhardwaj Purdue University Andrew Whittaker: Univ. of...

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

    Office of Environmental Management (EM)

    Jacobs Company, LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board...

  8. Hazard Categorization and Accident Analysis Techniques for Compliance...

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

    Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports by Diane Johnson he purpose of this DOE Standard is to...

  9. Freezing Spring Temperatures Damage Knobcone Pine

    E-Print Network [OSTI]

    Freezing Spring Temperatures Damage Knobcone Pine Stanley L. Krugman U. S. FOREST SERVICE RESEARCH, Stanley L. 1966. Freezing spring temperatures damage knobcone pine conelets. Berkeley, Calif.. Pacific pine, conelets, freezing temperature) Krugman, Stanley L. 1966. Freezing spring temperatures damage

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

    SciTech Connect (OSTI)

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

    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.

  11. 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.; Sandia National Labs., Albuquerque, NM )

    1989-08-01

    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.

  12. Material Selection for Accident Tolerant Fuel Cladding

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

    none,

    2014-07-01

    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

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

    SciTech Connect (OSTI)

    Bartholome, G.; Faidy, C.; Franco, C.

    1997-04-01

    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.

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

    E-Print Network [OSTI]

    Howe, B.; Scales, B.

    1997-01-01

    stream_source_info ESL-IE-97-04-27.pdf.txt stream_content_type text/plain stream_size 18135 Content-Encoding ISO-8859-1 stream_name ESL-IE-97-04-27.pdf.txt Content-Type text/plain; charset=ISO-8859-1 Beyond Leaks: Demand.... Adding sensors to detect when compressed air is required, and then automating how the compressed air is applied, can dramatically reduce both compressed air use and peak demand. COMPRESSED AIR DELIVERY AND CONTROL How compressed air is delivered...

  15. 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

    1997-04-01

    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.

  16. 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.

  17. 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

  18. 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 Antonio, TX); Laine, Edwin F. (Alamo, CA)

    1997-01-01

    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 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 though 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.

  19. 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)

    2001-01-01

    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.

  20. OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT INVESTIGATION REPORT

    E-Print Network [OSTI]

    Arnold, Elizabeth A.

    OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT INVESTIGATION REPORT James Madison University, Office of Risk Management, 131 West Grace Street, MSC 6703 Harrisonburg, VA 22807, Phone: 540-568-7812, Fax: 540/Injury: Student's Signature: Date: Return Original Form to: #12;OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT

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

    SciTech Connect (OSTI)

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

    2000-03-23

    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.

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

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    1999-10-20

    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.

  3. BWR containment failure analysis during degraded-core accidents

    SciTech Connect (OSTI)

    Yue, D.D.

    1982-06-06

    This paper presents a containment failure mode analysis during a spectrum of postulated degraded core accident sequences in a typical 1000-MW(e) boiling water reactor (BWR) with a Mark-I wetwell containment. Overtemperature failure of containment electric penetration assemblies (CEPAs) has been found to be the major failure mode during such accidents.

  4. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect (OSTI)

    1998-03-01

    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.

  5. Unusual refinery boiler tube failures due to corrosion by sulfuric acid induced by steam leaks

    SciTech Connect (OSTI)

    Lopez-Lopez, D.; Wong-Moreno, A.

    1998-12-31

    Corrosion by sulfuric acid in boilers is a low probability event because gas temperature and metal temperature of boiler tubes are high enough to avoid the condensation of sulfuric acid from flue gases. This degradation mechanism is frequently considered as an important cause of air preheaters materials degradation, where flue gases are cooled by heat transfer to the combustion air. Corrosion is associated to the presence of sulfuric acid, which condensates if metal temperature (or gas temperature) is below of the acid dew point. In economizer tubes, sulfuric acid corrosion is an unlikely event because flue gas and tube temperatures are normally over the acid dewpoint. In this paper, the failure analysis of generator tubes (similar to the economizer of bigger boilers) of two small oil-fired subcritical boilers is reported. It is concluded that sulfuric acid corrosion was the cause of the failure. The sulfuric acid condensation was due to the contact of flue gases containing SO{sub 3} with water-steam spray coming from leaks at the interface of rolled tube to the drum. Considering the information gathered from these two cases studied, an analysis of this failure mechanism is presented including a description of the thermodynamics condition of water leaking from the drum, and an analysis of the factors favoring it.

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

    Energy Savers [EERE]

    OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Development of Light Water Reactor Fuels with Enhanced Accident Tolerance Report to Congress April 2015 United States...

  7. BWR containment flooding during a large break LOCA under different core damage conditions

    SciTech Connect (OSTI)

    Alammar, M.A.; Trikouros, N.G.; Hansen, P.N. [GPU Nuclear Corp., Parsippany, NJ (United States)

    1996-07-01

    The BWR Owners` Group require containment flooding as part of their Severe Accident Management Guidelines. It is shown in this analysis that flooding the containment increases the risk to containment integrity unless it is accompanied by a venting strategy. Using a large recirculation pipe break scenario with delayed core spray initiation such that 30% of the core had melted and relocated to the lower head (TMI-2 accident core damage) a venting strategy is formulated such that containment pressure is kept within Emergency Operating Procedures Limits. The strategy is based on the following criteria: (1) Venting starts when fission products scrubbing from the drywell atmosphere is completed; (2) Venting periods should be short to present water discharge through the vent due to level; (3) External injection may need to be terminated during venting to reduce interference with pressure behavior. The scenario was run until the drywell and the reactor pressure vessel were flooded above the top of active fuel elevation.

  8. The detection of carbon dioxide leaks using quasi-tomographic laser absorption spectroscopy measurements in variable wind

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

    Levine, Z. H.; Pintar, A. L.; Dobler, J.; Blume, N.; Braun, M.; Zaccheo, T. S.; Pernini, T. G.

    2015-11-24

    Laser Absorption Spectroscopy (LAS) has been used over the last several decades for the measurement of trace gasses in the atmosphere. For over a decade, LAS measurements from multiple sources and tens of retroreflectors have been combined with sparse-sample tomography methods to estimate the 2-D distribution of trace gas concentrations and underlying fluxes from pointlike sources. In this work, we consider the ability of such a system to detect and estimate the position and rate of a single point leak which may arise as a failure mode for carbon dioxide storage. The leak is assumed to be at a constantmore »rate giving rise to a plume with a concentration and distribution that depend on the wind velocity. We demonstrate the ability of our approach to detect a leak using numerical simulation and a preliminary measurement.« less

  9. Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

    SciTech Connect (OSTI)

    Zhao, Haihua; Zhang, Hongbin; Zou, Ling; Anders, David; Martineau, Richard

    2014-10-01

    The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC) system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.

  10. MELCOR accident analysis for ARIES-ACT

    SciTech Connect (OSTI)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    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.

  11. Big Rock Point severe accident management strategies

    SciTech Connect (OSTI)

    Brogan, B.A. [Consumers Power Co., Charlevoix, MI (United States); Gabor, J.R. [Dames and Moore, Westmont, IL (United States)

    1996-07-01

    December 1994, the Nuclear Energy Institute (NEI) issued guidance relative to the formal industry position on Severe Accident Management (SAM) approved by the NEI Strategic Issues Advisory Committee on November 4, 1994. This paper summarizes how Big Rock Point (BRP) has and continues to address SAM strategies. The historical accounting portion of this presentation includes a description of how the following projects identified and defined the current Big Rock Point SAM strategies: the 1981 Level 3 Probabilistic Risk Assessment performance; the development of the Plant Specific Technical Guidelines from which the symptom oriented Emergency Operating Procedures (EOPs) were developed; the Control Room Design Review; and, the recent completion of the Individual Plant Evaluation (IPE). In addition to the historical presentation deliberation, this paper the present activities that continue to stress SAM strategies.

  12. Guidelines for accident prevention and emergency preparedness

    SciTech Connect (OSTI)

    Fthenakis, V.M.; Morris, S.C.; Moskowitz, P.D.

    1993-05-01

    This report reviews recent developments in the guidelines on chemical accident prevention, risk assessment, and management of chemical emergencies, principally in the United States and Europe, and discusses aspects of their application to developing countries. Such guidelines are either in the form of laws or regulations promulgated by governments, or of recommendations from international, professional, or non governmental organizations. In many cases, these guidelines specify lists of materials of concern and methods for evaluating safe usage of these materials and recommend areas of responsibility for different organizations; procedures to be included in planning, evaluation, and response; and appropriate levels of training for different classes of workers. Guidelines frequently address the right of communities to be informed of potential hazards and address ways for them to participate in planning and decision making.

  13. Criteria for calculating the efficiency of HEPA filters during and after design basis accidents

    SciTech Connect (OSTI)

    Bergman, W.; First, M.W.; Anderson, W.L.; Gilbert, H.; Jacox, J.W.

    1994-12-01

    We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data.

  14. 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

  15. Accident management then and now: Progress since Di Salvo's work

    SciTech Connect (OSTI)

    Shotkin, L.M.

    1991-01-01

    The nuclear industry is now initiating a serious effort to define the elements of an accident management program at each utility with an operating reactor, which is a significant change in conditions from those in 1985, when the work of Di Salvo et al. was published. Each utility is now conducting an individual plant examination (IPE) to uncover plant vulnerabilities to severe accidents. In conjunction with the IPE program, the Nuclear Utility Management and Resources committee, the Electric Power Research Institute, and owners' groups are developing an accident management program. This program is emphasizing the management program. This program is emphasizing the management of severe accidents (i.e., accidents that proceed to significant core melt) including strategies for managing ex-vessel events. Attention is also being paid to interfacing any severe accident management strategies with existing emergency operating procedures already in place at utilities. The industry program is addressing the five elements define by the US Nuclear Regulatory Commission (NRC): (1) strategies; (2) instrumentation; (3) guidance and computational aids; (4) organization and decision making; and (5) training. It will also be able to accept new information as it becomes available from ongoing efforts to better understand severe accidents and how to manage them effectively.

  16. Post-accident inhalation exposure and experience with plutonium

    SciTech Connect (OSTI)

    Shinn, J

    1998-06-01

    This paper addresses the issue of inhalation exposure immediately afterward and for a long time following a nuclear accident. For the cases where either a nuclear weapon burns or explodes prior to nuclear fission, or at locations close to a nuclear reactor accident containing fission products, a major concern is the inhalation of aerosolized plutonium (Pu) particles producing alpha-radiation. We have conducted field studies of Pu- contaminated real and simulated accident sites at Bikini, Johnston Atoll, Tonopah (Nevada), Palomares (Spain), Chernobyl, and Maralinga (Australia).

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

    E-Print Network [OSTI]

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

    1993-01-01

    Central steam is commonly distributed through direct burial lines protected by an outer conduit. These underground conduit systems are subject to electrolytic corrosion. Failure of the outer casing permits water intrusion and damage to insulation...

  18. Electromagnetic Interrogation Techniques Damage Detection

    E-Print Network [OSTI]

    Electromagnetic Interrogation Techniques for Damage Detection H. T. Banks and M. L. Joyner Center.P. Winfree Nasa Langley Research Center Hampton, VA Plenary Lecture, Electromagnetic Nondestructive Evaluation 2001 (ENDE 2001), Kobe, Japan, May 18-19, 20001 #12;Electromagnetic Interrogation Techniques

  19. Electromagnetic Interrogation Techniques Damage Detection

    E-Print Network [OSTI]

    Electromagnetic Interrogation Techniques for Damage Detection H. T. Banks #3; and M. L. Joyner Wincheski and W.P. Winfree Nasa Langley Research Center Hampton, VA #3; Plenary Lecture, Electromagnetic Nondestructive Evaluation 2001 (ENDE 2001), Kobe, Japan, May 18­19, 20001 #12; Electromagnetic Interrogation

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

    SciTech Connect (OSTI)

    Bouchard, P.J.

    1997-04-01

    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.

  1. Geochemical Impacts of Leaking CO2 from Subsurface Storage Reservoirs to Unconfined and Confined Aquifers

    SciTech Connect (OSTI)

    Qafoku, Nikolla; Brown, Christopher F.; Wang, Guohui; Sullivan, E. C.; Lawter, Amanda R.; Harvey, Omar R.; Bowden, Mark

    2013-04-15

    Experimental research work has been conducted and is undergoing at Pacific Northwest National Laboratory (PNNL) to address a variety of scientific issues related with the potential leaks of the carbon dioxide (CO2) gas from deep storage reservoirs. The main objectives of this work are as follows: • Develop a systematic understanding of how CO2 leakage is likely to influence pertinent geochemical processes (e.g., dissolution/precipitation, sorption/desorption and redox reactions) in the aquifer sediments. • Identify prevailing environmental conditions that would dictate one geochemical outcome over another. • Gather useful information to support site selection, risk assessment, policy-making, and public education efforts associated with geological carbon sequestration. In this report, we present results from experiments conducted at PNNL to address research issues related to the main objectives of this effort. A series of batch and column experiments and solid phase characterization studies (quantitative x-ray diffraction and wet chemical extractions with a concentrated acid) were conducted with representative rocks and sediments from an unconfined, oxidizing carbonate aquifer, i.e., Edwards aquifer in Texas, and a confined aquifer, i.e., the High Plains aquifer in Kansas. These materials were exposed to a CO2 gas stream simulating CO2 gas leaking scenarios, and changes in aqueous phase pH and chemical composition were measured in liquid and effluent samples collected at pre-determined experimental times. Additional research to be conducted during the current fiscal year will further validate these results and will address other important remaining issues. Results from these experimental efforts will provide valuable insights for the development of site-specific, generation III reduced order models. In addition, results will initially serve as input parameters during model calibration runs and, ultimately, will be used to test model predictive capability and competency. The results from these investigations will provide useful information to support site selection, risk assessment, and public education efforts associated with geological, deep subsurface CO2 storage and sequestration.

  2. 18 IEEE Transactions onPower Delivery, Vol. 14, No.1, January 1999 Leak Location in Fluid Filled Cables

    E-Print Network [OSTI]

    . Self-contained, low and medium- pressure, fluid filled cables which have an oil channel in the middle advantages over the conventional method of freeze and pressure testing. Description ofthe method and results, dielectric fluid leaks, fluid-filled cable. Introduction High Pressure Fluid Filled (HPFF), pipe-type cable

  3. Reducing Leaks in Cast Pump and Valve Bodies using Solidification Simulation Raymond Monroe, Steel Founders' Society of America

    E-Print Network [OSTI]

    Beckermann, Christoph

    shrinkage. This provides a way to assess casting quality prior to production at sensitivity levels the liquid and so the shrinkage associated with this transition must be managed through clever casting designReducing Leaks in Cast Pump and Valve Bodies using Solidification Simulation Raymond Monroe, Steel

  4. Blackbox Traceable CP-ABE: How to Catch People Leaking Their Keys by Selling Decryption Devices on eBay

    E-Print Network [OSTI]

    International Association for Cryptologic Research (IACR)

    Blackbox Traceable CP-ABE: How to Catch People Leaking Their Keys by Selling Decryption Devices@cityu.edu.hk Abstract. In the context of Ciphertext-Policy Attribute-Based Encryption (CP-ABE), if a decryption device with policies satisfied by SD, no one including the CP-ABE authorities can identify the malicious user(s) who

  5. Revised paper Leak NED 1997.doc 8:53 25.10.2002 1 Submitted to Nuclear Engineering and Design

    E-Print Network [OSTI]

    Cizelj, Leon

    Revised paper Leak NED 1997.doc 8:53 25.10.2002 1 Submitted to Nuclear Engineering and Design in nuclear power plants with pressurized water reactors comprise most of the reactor coolant pressure and fitness-for-service criteria for degraded steam generator tubes are being implemented on a worldwide basis

  6. Collateral damage: Evolution with displacement of fracture distribution and secondary fault strands in fault damage zones

    E-Print Network [OSTI]

    Savage, Heather M.; Brodsky, Emily E.

    2011-01-01

    E. McCallum (1999), Reservoir damage around faults: OutcropSkar (2005), Controls on damage zone asymmetry of a normal2007), The evolution of the damage zone with fault growth in

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

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

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

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

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

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

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

    Energy Savers [EERE]

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

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

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

    event and causal factor analysis. Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 More Documents & Publications DOE-HDBK-1092-1998...

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

    Energy Savers [EERE]

    Review, Savannah River National Laboratory - January 2012 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 Audit Report: OAS-L-06-15...

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

    E-Print Network [OSTI]

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

    2009-01-01

    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 ...

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

    Energy Savers [EERE]

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

  14. The Effect of Removing Accidents Repeaters From the Road

    E-Print Network [OSTI]

    Ferreira, Joseph Jr.

    In newspaper editorials, public commentaries and the like, licensing authorities are often advised to solve the "accident problem" by taking the "nut behind the wheel" off the road. This paper uses six-year driver records ...

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

    E-Print Network [OSTI]

    Higgins, Saoirse, 1966-

    2004-01-01

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

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

    E-Print Network [OSTI]

    Ferrari, Silvia

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

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

    Energy Savers [EERE]

    July 30, 2013, Electrical Fatality on the Bandon-Rogue No. 1 115kV Line at the Bonneville Power Administration Accident Investigation of the July 30, 2013, Electrical Fatality on...

  18. Some methods of estimating uncertainty in accident reconstruction

    E-Print Network [OSTI]

    Milan Batista

    2011-07-20

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

  19. Modeling control room crews for accident sequence analysis

    E-Print Network [OSTI]

    Huang, Y. (Yuhao)

    1991-01-01

    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 ...

  20. Type B Accident Investigation Board Report BNFL, Inc. Employee...

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

    17, 2003, at approximately 7:15 a.m., an accident occurred at the U.S. Department of Energy (DOE) East Tennessee Technology Park, Building K-31. An employee (Pipefitter) of...

  1. Method for producing damage resistant optics

    DOE Patents [OSTI]

    Hackel, Lloyd A. (Livermore, CA); Burnham, Alan K. (Livermore, CA); Penetrante, Bernardino M. (San Ramon, CA); Brusasco, Raymond M. (Livermore, CA); Wegner, Paul J. (Livermore, CA); Hrubesh, Lawrence W. (Pleasanton, CA); Kozlowski, Mark R. (Windsor, CA); Feit, Michael D. (Livermore, CA)

    2003-01-01

    The present invention provides a system that mitigates the growth of surface damage in an optic. Damage to the optic is minimally initiated. In an embodiment of the invention, damage sites in the optic are initiated, located, and then treated to stop the growth of the damage sites. The step of initiating damage sites in the optic includes a scan of the optic using a laser to initiate defects. The exact positions of the initiated sites are identified. A mitigation process is performed that locally or globally removes the cause of subsequent growth of the damaged sites.

  2. Material selection for accident tolerant fuel cladding

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

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    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. Therefore, 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.« less

  3. Study on drywell cooler applicability to severe accident management

    SciTech Connect (OSTI)

    Nakagawa, Takahiro [Information and manufacturing systems division, Toshiba Plant Systems and Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Akinaga, Makoto [Power and Industrial Systems R and D Center, Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki, 210-0862 (Japan); Hamazaki, Ryoichi [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Matsuo, Toshihiro [Nuclear Power Engineering Department, Tokyo Electric Power Company, 1-3 Uchisaiwai-cho 1-chome, Chiyoda-ku, Tokyo 100-0011 (Japan); Hashimoto, Kouji [Nuclear Plant Engineering Department, HITACHI, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-8511 (Japan)

    2004-07-01

    This paper concerns applicability of drywell cooler (DWC) heat removal under severe accident condition in BWR plants. Newly developed heat removal models based on DWC heat removal experiments were built into the MAAP3 code. And then, two types of Japanese BWR were selected to evaluate DWC heat removal performance under typical severe accident scenarios. According to the results of the evaluation, DWC delays or prevents containment failure or venting. (authors)

  4. Extension of emergency operating procedures for severe accident management

    SciTech Connect (OSTI)

    Chiang, S.C. (Taiwan Power Company, Taipei (Taiwan, Province of China))

    1992-01-01

    To enhance the capability of reactor operators to cope with the hypothetical severe accident its the key issue for utilities. Taiwan Power Company has started the enhancement programs on extension of emergency operating procedures (EOPs). It includes the review of existing LOPs based on the conclusions and recommendations of probabilistic risk assessment studies to confirm the operator actions. Then the plant specific analysis for accident management strategy will be performed and the existing EOPs will be updated accordingly.

  5. Percutaneous Transhepatic Biliary Drainage in the Management of Postsurgical Biliary Leaks in Patients with Nondilated Intrahepatic Bile Ducts

    SciTech Connect (OSTI)

    Cozzi, Guido, E-mail: guido.cozzi@istitutotumori.mi.it; Severini, Aldo; Civelli, Enrico; Milella, Marco [National Cancer Institute (Istituto Nazionale Tumori), Department of Radiology, Radiologia 3 Unit (Italy); Pulvirenti, Andrea [National Cancer Institute (Istituto Nazionale Tumori), Department of Surgery, Gastrointestinal Surgery and Liver Transplantation Unit (Italy); Salvetti, Monica [National Cancer Institute (Istituto Nazionale Tumori), Department of Radiology, Radiologia 3 Unit (Italy); Romito, Raffaele [National Cancer Institute (Istituto Nazionale Tumori), Department of Surgery, Gastrointestinal Surgery and Liver Transplantation Unit (Italy); Suman, Laura; Chiaraviglio, Francesca [National Cancer Institute (Istituto Nazionale Tumori), Department of Radiology, Radiologia 3 Unit (Italy); Mazzaferro, Vincenzo [National Cancer Institute (Istituto Nazionale Tumori), Department of Surgery, Gastrointestinal Surgery and Liver Transplantation Unit (Italy)

    2006-06-15

    Purpose. To assess the feasibility of percutaneous transhepatic biliary drainage (PTBD) for the treatment of postsurgical biliary leaks in patients with nondilated intrahepatic bile ducts, its efficacy in restoring the integrity of bile ducts, and technical procedures to reduce morbidity. Methods. Seventeen patients out of 936 undergoing PTBD over a 20-year period had a noncholestatic liver and were retrospectively reviewed. All patients underwent surgery for cancer and suffered a postsurgical biliary leak of 345 ml/day on average; 71% were in poor condition and required permanent nutritional support. An endoscopic approach failed or was excluded due to inaccessibility of the bile ducts. Results. Established biliary leaks and site of origin were diagnosed an average of 21 days (range 1-90 days) after surgery. In all cases percutaneous access to the biliary tree was achieved. An external (preleakage) drain was applied in 7 cases, 9 patients had an external-internal fistula bridging catheter, and 1 patient had a percutaneous hepatogastrostomy. Fistulas healed in an average of 31 days (range 3-118 days ) in 15 of 17 patients (88%) following PTBD. No major complications occurred after drainage. Post-PTBD cholangitis was observed in 6 of 17 patients (35%) and was related to biliary sludge formation occurring mostly when drainage lasted >30 days and was of the external-internal type. Median patient survival was 17.7 months and in all cases the repaired biliary leaks remained healed. Conclusions. PTBD is a feasible, effective, and safe procedure for the treatment of postsurgical biliary leaks. It is therefore a reliable alternative to surgical repair, which entails longer hospitalization and higher costs.

  6. AP1000{sup R} severe accident features and post-Fukushima considerations

    SciTech Connect (OSTI)

    Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

  7. Detection of damage in axial (membrane) systems

    SciTech Connect (OSTI)

    Duffey, T.A. [Duffey (T.A.), Tijeras, NM (United States); Baker, W.E. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Mechanical Engineering; Farrar, C.R. [Los Alamos National Lab., NM (United States); Rhee, W.H. [Texas Tech Univ., Lubbock, TX (United States). Dept. of Mechanical Engineering

    1998-12-31

    In a recent paper, two methods of damage identification (Modified Damage Index and Change-in-Flexibility) were applied to detection of damage in an 8-DOF vibrating system. The goal of the work was to detect damage (reduction in stiffness of one or more of the elements) as well as to locate the particular damaged elements (S). However, the investigation was limited to numerical simulations only. In this paper, a physical, spring-mass model of a similar, degenerate 8-DOF system (7 normal modes plus a rigid-body mode) was constructed. Experiments were then performed and the modal properties of the system were determined in undamaged and damaged states. Excitation was provided either by an impact hammer or by an electromechanical shaker. Damage was induced by replacing one of the springs with a spring of lower stiffness. The Modified Damage Index method clearly isolated the location of damage for a variety of damage locations and levels of damage. The Change-in-Flexibility method, however, was found to be less reliable. The ability of the method to locate damage depended strongly on location and the level of damage as well as the number of modes included.

  8. Type B Accident Investigation Board Report on the March 27, 1998, Rotating Shaft Accident at the Ames Laboratory, Ames, Iowa

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by John Kennedy, Acting Manager, Chicago Operations Office, U.S. Department of Energy (DOE).

  9. A damage model for fracking

    E-Print Network [OSTI]

    Norris, J Quinn; Rundle, John B

    2015-01-01

    Injections of large volumes of water into tight shale reservoirs allows the extraction of oil and gas not previously accessible. This large volume "super" fracking induces damage that allows the oil and/or gas to flow to an extraction well. The purpose of this paper is to provide a model for understanding super fracking. We assume that water is injected from a small spherical cavity into a homogeneous elastic medium. The high pressure of the injected water generates hoop stresses that reactivate natural fractures in the tight shales. These fractures migrate outward as water is added creating a spherical shell of damaged rock. The porosity associated with these fractures is equal to the water volume injected. We obtain an analytic expression for this volume. We apply our model to a typical tight shale reservoir and show that the predicted water volumes are in good agreement with the volumes used in super fracking.

  10. KU Public Safety Office Criminal Damage

    E-Print Network [OSTI]

    KU Public Safety Office ! Criminal Damage The unidentified white male pictured below is a suspect in the damage of a Coca-Cola vending machine in the Parking Services lobby at 1501 Irving Hill Drive and damage to Coca-Cola vending machines across the campus. Suspect Description: W/M, 5 feet 10 inches, 150

  11. Multiscale Modeling of Radiation Damage in

    E-Print Network [OSTI]

    Multiscale Modeling of Radiation Damage in Fusion Reactor Materials Brian D. Wirth, R.J. Kurtz-7405-Eng-48. #12;Presentation overview · Introduction to fusion reactor materials and radiation damage. tailor He HFIR isotopic tailor He HFIR target/RB He appmHe displacement damage (dpa) ffuussiioonn

  12. Density Functional Theory Models for Radiation Damage

    E-Print Network [OSTI]

    Density Functional Theory Models for Radiation Damage S.L. Dudarev EURATOM/CCFE Fusion Association and informative as the most advanced experimental techniques developed for the observation of radiation damage investigation and assessment of radiation damage effects, offering new insight into the origin of temperature

  13. Structural Damage Detection and Localization Using NETSHM

    E-Print Network [OSTI]

    Gnawali, Omprakash

    Structural Damage Detection and Localization Using NETSHM Krishna Chintalapudi, Jeongyeup Paek and localize damage in large civil structures. Structural engineers often implement and test SHM algorithms the intricacies of wireless networking, or the details of sensor data acquisition. We have implemented a damage

  14. DETECTION OF HISTORICAL PIPELINE LEAK PLUMES USING NON-INTRUSIVE SURFACE-BASED GEOPHYSICAL TECHNIQUES AT THE HANFORD NUCLEAR SITE WASHINGTON USA

    SciTech Connect (OSTI)

    SKORSKA MB; FINK JB; RUCKER DF; LEVITT MT

    2010-12-02

    Historical records from the Department of Energy Hanford Nuclear Reservation (in eastern WA) indicate that ruptures in buried waste transfer pipelines were common between the 1940s and 1980s, which resulted in unplanned releases (UPRs) of tank: waste at numerous locations. A number of methods are commercially available for the detection of active or recent leaks, however, there are no methods available for the detection of leaks that occurred many years ago. Over the decades, leaks from the Hanford pipelines were detected by visual observation of fluid on the surface, mass balance calculations (where flow volumes were monitored), and incidental encounters with waste during excavation or drilling. Since these detection methods for historic leaks are so limited in resolution and effectiveness, it is likely that a significant number of pipeline leaks have not been detected. Therefore, a technology was needed to detect the specific location of unknown pipeline leaks so that characterization technologies can be used to identify any risks to groundwater caused by waste released into the vadose zone. A proof-of-concept electromagnetic geophysical survey was conducted at an UPR in order to image a historical leak from a waste transfer pipeline. The survey was designed to test an innovative electromagnetic geophysical technique that could be used to rapidly map the extent of historical leaks from pipelines within the Hanford Site complex. This proof-of-concept test included comprehensive testing and analysis of the transient electromagnetic method (TEM) and made use of supporting and confirmatory geophysical methods including ground penetrating radar, magnetics, and electrical resistivity characterization (ERC). The results for this initial proof-of-concept test were successful and greatly exceeded the expectations of the project team by providing excellent discrimination of soils contaminated with leaked waste despite the interference from an electrically conductive pipe.

  15. Human factors review for Severe Accident Sequence Analysis (SASA)

    SciTech Connect (OSTI)

    Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

    1984-01-01

    The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure.

  16. Damage Detection in Plate Structures using Guided Ultrasonic Waves

    E-Print Network [OSTI]

    Jarmer, Gregory James Sylvester

    2009. “Evaluation of the Damage Detection Capability of alikelihood Estimation of Damage Location in Guided- waveStatistically-based Damage Detection in Geometrically-

  17. Micromechanical Damage Models for Continuous Fiber Reinforced Composite Materials

    E-Print Network [OSTI]

    Wu, Yi

    2013-01-01

    2008). Micromechanical modeling of damage and fracture ofmatrix viscoplasticity and evolving damage, Journal of theW.A. (1998). Stochastic damage evolution and failure in

  18. Blast damage mitigation of steel structures from near- contact charges

    E-Print Network [OSTI]

    Wolfson, Janet Crumrine

    2008-01-01

    OF CALIFORNIA, SAN DIEGO Blast Damage Mitigation of Steel35  Damage Levels Observed in LaboratoryFigure 3.34: Progression of damage for a Ballistic Loading

  19. Cognitive Empathy Following Orbitofrontal Cortex and Dorsolateral Prefrontal Cortex Damage

    E-Print Network [OSTI]

    Goodkind, Madeleine Shirley

    2010-01-01

    following bilateral damage to the human amygdala. Nature,as a measure of frontal lobe damage. Journal of Clinical andcaused by frontal damage fail to respond autonomically to

  20. RADIATION DAMAGE RESISTANCE OF REVERSE ELECTRODE GE COAXIAL DETECTORS

    E-Print Network [OSTI]

    Pehl, Richard H.

    2011-01-01

    Parker, "Radiation Damage of Germanium Detectors", Bull. Am.to radiation damage between the two detectors was clearlyRADIATION DAMAGE RESISTANCE OF REVERSE ELECTRODE GE COAXIAL DETECTORS

  1. Nonlocal Damage Models 8.1 Basic Types of Nonlocal Damage Formulations

    E-Print Network [OSTI]

    Jirasek, Milan

    Chapter 8 Nonlocal Damage Models 8.1 Basic Types of Nonlocal Damage Formulations 8.1.1 Formulations Motivated by Isotropic Damage A number of nonlocal concepts giving local response in the linear elastic damage model from Section 5.2. Certain models use a formulation in which the role of the equivalent

  2. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect (OSTI)

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K.

    2012-07-01

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  3. Accident Sequence Evaluation Program: Human reliability analysis procedure

    SciTech Connect (OSTI)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  4. Emission factors for leaks in refinery components in heavy liquid service

    SciTech Connect (OSTI)

    Taback, H.; Godec, M.

    1996-12-31

    The objective of this program was to provide sufficient screening data so that EPA can develop an official set of emission factors (expressed in lb/hr/component) for refinery components (valves, flanged connectors, non-flanged connectors, pumps, open-ended lines, and other) in heavy liquid (BL) service. To accomplish this, 211,000 existing HL screening values from Southern California refineries were compiled and compared with 2,500 new HL screening measurements taken at two refineries in the state of Washington. Since Southern California is an area in extreme non-attainment of the National Ambient Air Quality Standards (NAAQS) and therefore has tight emission control regulations, it was felt that its screening data may not be representative of refineries without tight emission controls. Thus, the Southern California screening data were compared to screening measurements at refineries in an area that is in attainment of the NAAQS and without emissions control, which is the case for those refineries in Washington. It was found that statistically there was no significant difference in emission factors between the two areas and, therefore, there appears to be no difference in emissions from heavy liquid components in areas with and without leak detection and repair (LDAR) programs. The new emission factors range from 1/7 to 1/3 times the current EPA emission factors. This program was sponsored by the American Petroleum Institute (API) and an API report will soon be released providing complete details.

  5. A Facility for Accurate Heat Load and Mass Leak Measurements on Superfluid Helium Valves

    E-Print Network [OSTI]

    Bézaguet, Alain-Arthur; Ferlin, G; Losserand-Madoux, R; Perin, A; Vandoni, Giovanna; Van Weelderen, R

    1999-01-01

    The superconducting magnets of the Large Hadron Collider (LHC) will be protected by safety relief valves operating at 1.9 K in superfluid helium (HeII). A test facility was developed to precisely determine the heat load and the mass leakage of cryogenic valves with HeII at their inlet. The temperature of the valve inlet can be varied from 1.8 K to 2 K for pressures up to 3.5 bar. The valve outlet pipe temperature can be regulated between 5 K and 20 K. The heat flow is measured with high precision using a Kapitza-resistance heatmeter and is also crosschecked by a vaporization measurement. After calibration, a precision of 10 mW for heat flows up to 1.1 W has been achieved. The helium leak can be measured up to 15 mg/s with an accuracy of 0.2 mg/s. We present a detailed description of the test facility and the measurements showing its performances.

  6. AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION

    SciTech Connect (OSTI)

    Jerry Myers

    2003-11-12

    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. This second six-month technical report summarizes the progress made towards defining, designing, and developing the hardware and software segments of the airborne, optical remote methane and ethane sensor. The most challenging task to date has been to identify a vendor capable of designing and developing a light source with the appropriate output wavelength and power. This report will document the work that has been done to identify design requirements, and potential vendors for the light source. Significant progress has also been made in characterizing the amount of light return available from a remote target at various distances from the light source. A great deal of time has been spent conducting laboratory and long-optical path target reflectance measurements. This is important since it helps to establish the overall optical output requirements for the sensor. It also reduces the relative uncertainty and risk associated with developing a custom light source. The data gathered from the optical path testing has been translated to the airborne transceiver design in such areas as: fiber coupling, optical detector selection, gas filters, and software analysis. Ophir will next, summarize the design progress of the transceiver hardware and software development. Finally, Ophir will discuss remaining project issues that may impact the success of the project.

  7. AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPLINE LEAK DETECTION

    SciTech Connect (OSTI)

    Jerry Myers

    2004-05-12

    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 third six-month technical report contains a summary of the progress made towards finalizing the design and assembling the airborne, remote methane and ethane sensor. The vendor has been chosen and is on contract to develop the light source with the appropriate linewidth and spectral shape to best utilize the Ophir gas correlation software. Ophir has expanded upon the target reflectance testing begun in the previous performance period by replacing the experimental receiving optics with the proposed airborne large aperture telescope, which is theoretically capable of capturing many times more signal return. The data gathered from these tests has shown the importance of optimizing the fiber optic receiving fiber to the receiving optic and has helped Ophir to optimize the design of the gas cells and narrowband optical filters. Finally, Ophir will discuss remaining project issues that may impact the success of the project.

  8. 340 Facility Secondary Containment and Leak Detection Project W-302 Functional Design Criteria

    SciTech Connect (OSTI)

    Stordeur, R.T.

    1995-03-01

    This functional design criteria for the upgrade to the 340 radioactive liquid waste storage facility (Project W-302) specifically addresses the secondary containment issues at the current vault facility of the 340 Complex. This vault serves as the terminus for the Radioactive Liquid Waste System (RLWS). Project W-302 is necessary in order to bring this portion of the Complex into full regulatory compliance. The project title, ``340 Facility Secondary Containment and Leak Detection``, illustrates preliminary thoughts of taking corrective action directly upon the existing vault (such as removing the tanks, lining the vault, and replacing tanks). However, based on the conclusion of the engineering study, ``Engineering Study of the 300 Area Process Wastewater Handling System``, WHC-SD-WM-ER-277 (as well as numerous follow-up meetings with cognizant staff), this FDC prescribes a complete replacement of the current tank/vault system. This offers a greater array of tanks, and provides greater operating flexibility and ease of maintenance. This approach also minimizes disruption to RLWS services during ``tie-in``, as compared to the alternative of trying to renovate the old vault. The proposed site is within the current Complex area, and maintains the receipt of RLWS solutions through gravity flow.

  9. The criteria of fracture in the case of the leak of pressure vessels

    SciTech Connect (OSTI)

    Habil; Ziliukas, A.

    1997-04-01

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  10. Remarks on numerical simulation of the PKN model of hydrofracturing in proper variables. Various leak-off regimes

    E-Print Network [OSTI]

    Kusmierczyk, P; Wrobel, M

    2012-01-01

    The problem of hydraulic fracture for the PKN model is considered within the framework of approach presented recently by Linkov (2011). The modified formulation is further enhanced by employing an improved regularized boundary condition near the crack tip. This increases solution accuracy especially for singular leak-off regimes. A new dependent variable having clear physical sense is introduced. A comprehensive analysis of numerical algorithms based on various dependent variables is provided.

  11. The {open_quotes}leak-before-break{close_quotes} applicability in decision support system {open_quotes}strength{close_quotes}

    SciTech Connect (OSTI)

    Torop, V.M.; Orynyak, I.V. [Institute for Problems of Strength, Kiev (Ukraine); Kutovoy, O.L. [Institute of Structure Integrity, Kiev (Ukraine)

    1997-04-01

    A software decision support system, STRENGTH, for application of leak before break analysis, is described. The background methodology and sample application are outlined. The program allows multioptional computation of loading parameters for different types of defects, and variable properties for metals and welded joints. Structural strength is assessed, and service life predictions are made. The program is used to analyze specific defects identified by nondestructive testing.

  12. A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors

    SciTech Connect (OSTI)

    S. Khericha

    2011-06-01

    This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.

  13. Effect of shape reactivity on the rod-ejection accident

    SciTech Connect (OSTI)

    Neogy, P.; Carew, J.F.

    1982-09-01

    The shape reactivity has a significant influence on the rod ejection accident. After the control rod is fully ejected from the core, the neutron flux undergoes a large reduction at the ejected rod location. The corresponding effect on the control reactivity is comparable in magnitude to the Doppler reactivity, and makes a significant contribution to limiting the power excursion during the transient. The neglect of this effect in point kinetics and space time synthesis analyses of the rod ejection accident may account in part for the large degree of conservatism usually associated with these analyses.

  14. A Perspective on Long-Term Recovery Following the Fukushima Nuclear Accident - 12075

    SciTech Connect (OSTI)

    Chen, S.Y. [Environmental Science Division, Argonne National Laboratory, Argonne, IL (United States)

    2012-07-01

    The tragic events at the Fukushima Daiichi Nuclear Power Station began occurring on March 11, 2011, following Japan's unprecedented earthquake and tsunami. The subsequent loss of external power and on-site cooling capacity severely compromised the plant's safety systems, and subsequently, led to core melt in the affected reactors and damage to spent nuclear fuel in the storage pools. Together with hydrogen explosions, this resulted in a substantial release of radioactive material to the environment (mostly Iodine-131 and Cesium- 137), prompting an extensive evacuation effort. The latest release estimate places the event at the highest severity level (Level 7) on the International Nuclear Event Scale, the same as the Chernobyl accident of 1986. As the utility owner endeavored to stabilize the damaged facility, environmental contamination continued to propagate and affect every aspect of daily life in the affected region of Japan. Elevated levels of radioactivity (mostly dominated by Cs-137 with the passage of time) were found in soil, drinking water, vegetation, produce, seafood, and other foodstuffs. An estimated 80,000 to 90,000 people were evacuated; more evacuations are being contemplated months after the accident, and a vast amount of land has become contaminated. Early actions were taken to ban the shipment and sale of contaminated food and drinking water, followed by later actions to ban the shipment and sale of contaminated beef, mushrooms, and seafood. As the event continues to evolve toward stabilization, the long-term recovery effort needs to commence - a process that doubtless will involve rather complex decision-making interactions between various stakeholders. Key issues that may be encountered and considered in such a process include (1) socio-political factors, (2) local economic considerations, (3) land use options, (4) remediation approaches, (5) decontamination methods, (6) radioactive waste management, (7) cleanup levels and options, and (8) government policies, among others. This paper offers a perspective on this likely long and arduous journey toward establishing a 'new normal' that will ultimately take shape. Toward this end, it is important to evaluate the 'optimization' process advocated by the international community in achieving long-term recovery from this particularly fateful event in Fukushima. In the process, experience and lessons learned from past events will be fully evaluated and considered. (author)

  15. Nanofoams Response to Radiation Damage

    SciTech Connect (OSTI)

    Fu, Engang [Los Alamos National Laboratory; Serrano De Caro, Magdalena [Los Alamos National Laboratory; Wang, Yongqiang [Los Alamos National Laboratory; Nastasi, Michael [Nebraska Center for Energy Sciences Research, University of Nebraska-Lincoln, NE 68508; Zepeda-Ruiz, Luis [PLS, Lawrence Livermore National Laboratory, Livermore, CA 94551; Bringa, Eduardo M. [CONICET and Inst. Ciencias Basicas, Universidad Nacional de Cuyo, Mendoza, 5500 Argentina; Baldwin, Jon K. [Los Alamos National Laboratory; Caro, Jose A. [Los Alamos National Laboratory

    2012-07-30

    Conclusions of this presentation are: (1) np-Au foams were successfully synthesized by de-alloying process; (2) np-Au foams remain porous structure after Ne ion irradiation to 1 dpa; (3) SFTs were observed in irradiated np-Au foams with highest and intermediate flux, while no SFTs were observed with lowest flux; (4) SFTs were observed in irradiated np-Au foams at RT, whereas no SFTs were observed at LNT irradiation; (5) The diffusivity of vacancies in Au at RT is high enough so that the vacancies have enough time to agglomerate and thus collapse. As a result, SFTs were formed; (6) The high flux created much more damage/time, vacancies don't have enough time to diffuse or recombine. As a result, SFTs were formed.

  16. Type A Accident Investigation Board Report on the July 1, 2008...

    Office of Environmental Management (EM)

    July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration Type A Accident Investigation Board Report on the July 1, 2008, of the Vehicle Fatality...

  17. Type A Accident Investigation Board Report on the April 3, 1995...

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

    1995 The accident under investigation occurred on April 3, 1995, at approximately 10:46 a.m. As a result of the accident, a Wackenhut Services, Incorporated-Savannah River Site...

  18. The Effect of the 18-Year Old Drinking Age on Auto Accidents

    E-Print Network [OSTI]

    Cucchiaro, Stephen

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

  19. Accident Analysis and Prevention 42 (2010) 364371 Contents lists available at ScienceDirect

    E-Print Network [OSTI]

    Boggess, May M.

    2010-01-01

    number of studies reported age-specific accident rates for heavy vehicles for the spec- trum of driver' shaped curve indicates a higher risk of accident involvement for both younger and older drivers. More

  20. Utilizing an encroachment probability benefit-cost model to estimate accident reduction factors 

    E-Print Network [OSTI]

    Hayes, Carolyn A

    1997-01-01

    Improving safety on Texas roadways is a major public concern. Over the years, the Texas Department of Transportation and other highway agencies have become interested in reducing society's accident cost while maximizing returns on accident...

  1. Three dimensional effects in analysis of PWR steam line break accident

    E-Print Network [OSTI]

    Tsai, Chon-Kwo

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

  2. AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION

    SciTech Connect (OSTI)

    Jerry Myers

    2003-05-13

    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. This six-month technical report summarizes the progress for each of the proposed tasks, discusses project concerns, and outlines near-term goals. Ophir has completed a data survey of two major natural gas pipeline companies on the design requirements for an airborne, optical remote sensor. The results of this survey are disclosed in this report. A substantial amount of time was spent on modeling the expected optical signal at the receiver at different absorption wavelengths, and determining the impact of noise sources such as solar background, signal shot noise, and electronic noise on methane and ethane gas detection. Based upon the signal to noise modeling and industry input, Ophir finalized the design requirements for the airborne sensor, and released the critical sensor light source design requirements to qualified vendors. Responses from the vendors indicated that the light source was not commercially available, and will require a research and development effort to produce. Three vendors have responded positively with proposed design solutions. Ophir has decided to conduct short path optical laboratory experiments to verify the existence of methane and absorption at the specified wavelength, prior to proceeding with the light source selection. Techniques to eliminate common mode noise were also evaluated during the laboratory tests. Finally, Ophir has included a summary of the potential concerns for project success and has established future goals.

  3. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    SciTech Connect (OSTI)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  4. Corrosion-induced damage raises serious implications

    SciTech Connect (OSTI)

    Kane, R.D.; Cayard, M.S. [CLI International, Inc., Houston, TX (United States)

    1997-06-01

    One of the most difficult and often underestimated aspects of pipeline rehabilitation is the assessment of corrosion-induced damage. This question involves evaluation of damage from prior service as well as consideration of conditions which may pose additional time-dependent degradation which could affect the future serviceability of the pipeline. The present study examines the assessment of pipeline damage and rehabilitation requirements through knowledge of materials of construction, operating conditions, field inspection and service records.

  5. Method for assaying clustered DNA damages

    DOE Patents [OSTI]

    Sutherland, Betsy M.

    2004-09-07

    Disclosed is a method for detecting and quantifying clustered damages in DNA. In this method, a first aliquot of the DNA to be tested for clustered damages with one or more lesion-specific cleaving reagents under conditions appropriate for cleavage of the DNA to produce single-strand nicks in the DNA at sites of damage lesions. The number average molecular length (Ln) of double stranded DNA is then quantitatively determined for the treated DNA. The number average molecular length (Ln) of double stranded DNA is also quantitatively determined for a second, untreated aliquot of the DNA. The frequency of clustered damages (.PHI..sub.c) in the DNA is then calculated.

  6. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  7. Relating geometric design consistency and accident experience on two-lane rural highways 

    E-Print Network [OSTI]

    Glascock, Stephen Wade

    1991-01-01

    to provide more efficient transportation on the one hand, and this same compulsion pushes man to work and devise methods of reducing the probability and severity of accidents. For so long as we have humans at the control of the transport vehicle..., the driver, and the vehicle contribute to the occurrence and severity of accidents. Seldom is information included in accident analyses that addresses all of these factors. Even with more complete accident data bases, however, researchers often are unable...

  8. An analysis to determine correlations of freeway traffic accidents with specific geometric design features 

    E-Print Network [OSTI]

    Smith, Frank Miller

    1960-01-01

    was followed to select an accident frequency index was the development and evaluation of several experimental indices. Factors considered, Measurement of accident exposure based on volumes and the number of accidents which occurred served as the basic... of these conditions, the length of the through lanes, in each direction from the ramp terminal, over which accidents were counted was a variable which was considered for several indices. In the application of this factor, through-lane acci- dents were grouped...

  9. Detailed Analysis of In-Vessel Melt Progression in the Loss of Coolant Accident of OPR1000

    SciTech Connect (OSTI)

    Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2006-07-01

    An in-vessel severe accident progression has been analyzed to generate the basic data for an evaluation of the in-vessel severe accident management strategies and to identify the thermal hydraulic condition of the reactor vessel and the damage state of the in-vessel materials at a reactor vessel failure by using the SCDAP/RELAP5/MOD3.3 computer code during the Loss Of Coolant Accident (LOCA) without the Safety Injection (SI) of the OPR (Optimized Pressurize Reactor) 1000. Best estimate calculation of the small break LOCAs of 1.35 inch and 2 inch, the medium break LOCAs of 3 inch and a 4.28 inch, and a large break LOCA of 9.8 inch without the SI have been performed from a transient initiation to a reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that in all the transients, approximately 30-40 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of a reactor vessel failure. In the small and large break LOCAs, the reactor vessel failed at an early time of approximately 70-110 minutes after the transients were initiated. Since the Safety Injection Tanks (SITs) were actuated effectively in the medium break LOCAs, the reactor vessel failed at a later time of approximately 200-400 minutes after the transients were initiated. At the time of a reactor vessel failure, approximately 45-55 % of the fuel rod cladding was oxidized in the small and medium break LOCAs. However, approximately 20 % of the fuel rod cladding was oxidized because of a coolant loss through the break in the large break LOCA of the OPR1000. (authors)

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

    E-Print Network [OSTI]

    Long, Nicholas

    . It is designed to help reduce injuries by reducing unsafe or hazardous conditions and discouraging accident causing unsafe acts or practices. It applies to all SFASU employees and campus locations conditions to your supervisor and the Safety Department by filling out a Report of Safety or Health Hazard

  11. An analysis of accident experience at entrance ramps within construction work zones at long-term freeway reconstruction projects in Texas 

    E-Print Network [OSTI]

    Casteel, David Bryan

    1991-01-01

    , severe accidents, daytime accidents, and multi-vehicle accidents (other than rear-end accidents) increased disproportionately in entrance ramp areas during construction. Conversely, accident frequencies did not increase significantly (a =0. 05... in Virginia, found that accidents in construction zones during 1977 were less severe than normal highway accidents. A reported 35 percent of work zone accidents were rear-end collisions. This may have contributed to the decrease in severity of reported work...

  12. Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident into the Pacific Ocean after the Fukushima accident and analysis of its dispersion in Japanese coastal waters, J into the ocean from the Fukushima Daiichi nuclear power plant (NPP) after the accident in March 2011 and to gain

  13. Why System Safety Professionals Should Read Accident Reports C. M. Holloway*, C. W. Johnson

    E-Print Network [OSTI]

    Johnson, Chris

    Why System Safety Professionals Should Read Accident Reports C. M. Holloway*, C. W. Johnson *NASA, who regularly read accident reports reap important benefits. These benefits include an improved accident reports regularly. This is a shame. People from many different disciplines have much to gain

  14. Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment

    E-Print Network [OSTI]

    Del Moral , Pierre

    Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment #12;The SYSTEMS WITH APPLICATIONS TO ACCIDENT RISK ASSESSMENT DISSERTATION to obtain the doctor's degree promotor Prof. dr. A. Bagchi #12;Contents 1 Introduction 3 1.1 Accident risk assessment

  15. Prediction of Damage Zone Growth in Composites Using Continuum Damage Mechanics 

    E-Print Network [OSTI]

    McLendon, Wesley R.

    2010-07-14

    The continuum damage mechanics (CDM) approach is widely used to model damage in polymer matrix composite materials which are represented using the homogenized properties of the fiber and matrix constituents. CDM simplifies the problem of accounting...

  16. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    SciTech Connect (OSTI)

    Robb, Kevin R

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  17. Damage Identification with Linear Discriminant Operators

    SciTech Connect (OSTI)

    Farrar, C.R.; Nix, D.A.; Duffey, T.A.; Cornwell, P.J.; Pardoen, G.C.

    1999-02-08

    This paper explores the application of statistical pattern recognition and machine learning techniques to vibration-based damage detection. First, the damage detection process is described in terms of a problem in statistical pattern recognition. Next, a specific example of a statistical-pattern-recognition-based damage detection process using a linear discriminant operator, ''Fisher's Discriminant'', is applied to the problem of identifying structural damage in a physical system. Accelerometer time histories are recorded from sensors attached to the system as that system is excited using a measured input. Linear Prediction Coding (LPC) coefficients are utilized to convert the accelerometer time-series data into multi-dimensional samples representing the resonances of the system during a brief segment of the time series. Fisher's discriminant is then used to find the linear projection of the LPC data distributions that best separates data from undamaged and damaged systems. The method i s applied to data from concrete bridge columns as the columns are progressively damaged. For this case, the method captures a clear distinction between undamaged and damaged vibration profiles. Further, the method assigns a probability of damage that can be used to rank systems in order of priority for inspection.

  18. MODELING LONGITUDINAL DAMAGE IN SHIP COLLISIONS

    E-Print Network [OSTI]

    Brown, Alan

    . Performing Organization Name and Address Department of Aerospace and Ocean Engineering. 10. Work Unit No made excellent progress towards predicting damage penetration in ship collisions. This project focuses collision data for penetrating collisions. 17. Key Words ship collisions, longitudinal ship damage 18

  19. SURFACE GEOPHYSICAL EXPLORATION DEVELOPING NONINVASIVE TOOLS TO MONITOR PAST LEAKS AROUND HANFORD TANK FARMS

    SciTech Connect (OSTI)

    MYERS DA; RUCKER DF; LEVITT MT; CUBBAGE B; NOONAN GE; MCNEILL M; HENDERSON C

    2011-06-17

    A characterization program has been developed at Hanford to image past leaks in and around the underground storage tank facilities. The program is based on electrical resistivity, a geophysical technique that maps the distribution of electrical properties of the subsurface. The method was shown to be immediately successful in open areas devoid of underground metallic infrastructure, due to the large contrast in material properties between the highly saline waste and the dry sandy host environment. The results in these areas, confirmed by a limited number of boreholes, demonstrate a tendency for the lateral extent of the underground waste plume to remain within the approximate footprint of the disposal facility. In infrastructure-rich areas, such as tank farms, the conventional application of electrical resistivity using small point-source surface electrodes initially presented a challenge for the resistivity method. The method was then adapted to directly use the buried infrastructure as electrodes for both transmission of electrical current and measurements of voltage. For example, steel-cased wells that surround the tanks were used as long electrodes, which helped to avoid much of the infrastructure problems. Overcoming the drawbacks of the long electrode method has been the focus of our work over the past seven years. The drawbacks include low vertical resolution and limited lateral coverage. The lateral coverage issue has been improved by supplementing the long electrodes with surface electrodes in areas devoid of infrastructure. The vertical resolution has been increased by developing borehole electrode arrays that can fit within the small-diameter drive casing of a direct push rig. The evolution of the program has led to some exceptional advances in the application of geophysical methods, including logistical deployment of the technology in hazardous areas, development of parallel processing resistivity inversion algorithms, and adapting the processing tools to accommodate electrodes of all shapes and locations. The program is accompanied by a full set of quality assurance procedures that cover the layout of sensors, measurement strategies, and software enhancements while insuring the integrity of stored data. The data have been shown to be useful in identifying previously unknown contaminant sources and defining the footprint of precipitation recharge barriers to retard the movement of existing contamination.

  20. Type B Accident Investigation Board Report of the January 20, 1998, Electrical Accident at the Casa Grande Substation,South of Phoenix, Arizona

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type-B Accident Investigation Board appointed by Michael S.Cowan, Chief Program Officer, Western Area Power Administration.

  1. An accident analysis of the physical plant of the Agricultural and Mechanical College of Texas 

    E-Print Network [OSTI]

    Allen, Gary James

    1963-01-01

    to reduce the fre- quency and/or severity of future accidents. 2 I. THE PROBLEM Statement of the problem. The problem was to make an analysis of the accident records of the Physical Plant Department National Safety Council, Accident Prevention Manual..., based on the analysis of rollected data, as to what type and where safety coz zectiors were most needed to reduce the frequency az d/or severity of accidents. Significance of the problem. The importance of accident prevention was probably best...

  2. Application of the leak-before-break concept to the primary circuit piping of the Leningrad NPP

    SciTech Connect (OSTI)

    Eperin, A.P.; Zakharzhevsky, Yu.O.; Arzhaev, A.I. [and others

    1997-04-01

    A two-year Finnish-Russian cooperation program has been initiated in 1995 to demonstrate the applicability of the leak-before-break concept (LBB) to the primary circuit piping of the Leningrad NPP. The program includes J-R curve testing of authentic pipe materials at full operating temperature, screening and computational LBB analyses complying with the USNRC Standard Review Plan 3.6.3, and exchange of LBB-related information with emphasis on NDE. Domestic computer codes are mainly used, and all tests and analyses are independently carried out by each party. The results are believed to apply generally to RBMK type plants of the first generation.

  3. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect (OSTI)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

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

  4. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect (OSTI)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

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

    SciTech Connect (OSTI)

    1991-09-01

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

  6. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect (OSTI)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tesmer, Joseph R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  7. TABLE OF CONTENTS Accident Prevention Signs, Tags, Labels, Signals,

    E-Print Network [OSTI]

    US Army Corps of Engineers

    to meet or exceed ANSI and/or OSHA requirements. USACE facilities shall use signs based upon and contractors may opt to use signs meeting either the OSHA or ANSI standards for temporary use during the life.200; Accident Prevention Signs and Tags; e. ANSI/IEEE C95.2; f. ANSI Z136.1; g. ANSI Z535.1; h. ANSI Z535.2; i

  8. Phase II Accident Investigation Board Briefing | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy Bills andOrder 422.1, CONDUCTCritical Materials UsePhase II Accident

  9. War damages and reconstruction of Peruca dam

    SciTech Connect (OSTI)

    Nonveiller, E.; Rupcic, J.; Sever, Z.

    1999-04-01

    The paper describes the heavy damages caused by blasting in the Peruca rockfill dam in Croatia in January 1993. Complete collapse of the dam by overtopping was prevented through quick action of the dam owner by dumping clayey gravel on the lowest sections of the dam crest and opening the bottom outlet of the reservoir, thus efficiently lowering the water level. After the damages were sufficiently established and alternatives for restoration of the dam were evaluated, it was decided to construct a diaphragm wall through the damaged core in the central dam part as the impermeable dam element and to rebuild the central clay core at the dam abutments. Reconstruction works are described.

  10. Results of Tank-Leak Detection Demonstration Using Geophysical Techniques at the Hanford Mock Tank Site-Fiscal Year 2001

    SciTech Connect (OSTI)

    Barnett, D BRENT.; Gee, Glendon W.; Sweeney, Mark D.

    2002-03-01

    During July and August of 2001, Pacific Northwest National Laboratory (PNNL), hosted researchers from Lawrence Livermore and Lawrence Berkeley National laboratories, and a private contractor, HydroGEOPHYSICS, Inc., for deployment of the following five geophysical leak-detection technologies at the Hanford Site Mock Tank in a Tank Leak Detection Demonstration (TLDD): (1) Electrical Resistivity Tomography (ERT); (2) Cross-Borehole Electromagnetic Induction (CEMI); (3) High-Resolution Resistivity (HRR); (4) Cross-Borehole Radar (XBR); and (5) Cross-Borehole Seismic Tomography (XBS). Under a ''Tri-party Agreement'' with Federal and state regulators, the U.S. Department of Energy will remove wastes from single-shell tanks (SSTs) and other miscellaneous underground tanks for storage in the double-shell tank system. Waste retrieval methods are being considered that use very little, if any, liquid to dislodge, mobilize, and remove the wastes. As additional assurance of protection of the vadose zone beneath the SSTs, tank wastes and tank conditions may be aggressively monitored during retrieval operations by methods that are deployed outside the SSTs in the vadose zone.

  11. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

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

    2004-05-15

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

  12. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    SciTech Connect (OSTI)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  13. DAMAGE ASSESSMENT OF COMPOSITE PLATE STRUCTURES WITH UNCERTAINTY

    E-Print Network [OSTI]

    Boyer, Edmond

    DAMAGE ASSESSMENT OF COMPOSITE PLATE STRUCTURES WITH UNCERTAINTY Chandrashekhar M.* , Ranjan Uncertainties associated with a structural model and measured vibration data may lead to unreliable damage that material uncertainties in composite structures cause considerable problem in damage assessment which can

  14. New navel orangeworm sanitation standards ?could reduce almond damage

    E-Print Network [OSTI]

    Higbee, Bradley S.; Siegel, Joel P

    2009-01-01

    disruption, dispersal and damage prediction. Proc 34thtype and amount of insect damage. J Ag Food Chem 49:4513–9.standards could reduce almond damage by Bradley S. Higbee

  15. RADIATION DAMAGE TO BSCCO-2223 FROM 50 MEV PROTONS

    E-Print Network [OSTI]

    Zeller, A.F.; Ronningen, R.M.; Godeke, A.; Heilbronn, L.H.; McMahan-Norris, P.; Gupta, R.

    2007-01-01

    RADIATION DAMAGE TO BSCCO-2223 FROM 50 MEV PROTONS A. F.BSCCO-2223. Radiation damage. INTRODUCTION The magnets incomponents be resistant to damage. One solution [1] is to

  16. A Damage-Revelation Rationale for Coupon Remedies

    E-Print Network [OSTI]

    Polinsky, A. Mitchell; Rubinfeld, Daniel L.

    2006-01-01

    Bargaining and the Design of Damage Awards,” 10 Journal ofpage 1 Revised: March 2006 A DAMAGE-REVELATION RATIONALE FORin a setting in which damages vary among plaintiffs and are

  17. INSTANTANEOUS DAMAGE IDENTIFICATION AND LOCALIZATION THROUGH SPARSE LASER ULTRASONIC SCANNING

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    INSTANTANEOUS DAMAGE IDENTIFICATION AND LOCALIZATION THROUGH SPARSE LASER ULTRASONIC SCANNING This study proposes an instantaneous damage identification and localization technique through sparse laser ultrasonic signals are obtained, a damage index (DI) representing the violation of the linear reciprocity

  18. A Damage-Revelation Rationale for Coupon Remedies

    E-Print Network [OSTI]

    Polinsky, A. Mitchell; Rubinfeld, Daniel L

    2007-01-01

    Bargaining and the Design of Damage Awards,’’ 10 Journal ofGramlich, Fred. 1986. ‘‘Scrip Damages in Antitrust Cases,’’in the Assessment of Damages,’’ 39 Journal of Law and

  19. Assessing United States hurricane damage under different environmental conditions

    E-Print Network [OSTI]

    Maheras, Anastasia Francis

    2012-01-01

    Hurricane activity between 1979 and 2011 was studied to determine damage statistics under different environmental conditions. Hurricanes cause billions of dollars of damage every year in the United States, but damage ...

  20. Ubiquitylation, neddylation and the DNA damage response

    E-Print Network [OSTI]

    Brown, Jessica S.; Jackson, Stephen P.

    2015-04-01

    , collectively termed the DNA damage response (DDR), requires the recruitment and subsequent post-translational modification (PTM) of a complex network of proteins. Ubiquitin and the ubiquitin-like protein (UBL) SUMO have established roles in regulating...

  1. Thin Film Femtosecond Laser Damage Competition

    SciTech Connect (OSTI)

    Stolz, C J; Ristau, D; Turowski, M; Blaschke, H

    2009-11-14

    In order to determine the current status of thin film laser resistance within the private, academic, and government sectors, a damage competition was started at the 2008 Boulder Damage Symposium. This damage competition allows a direct comparison of the current state of the art of high laser resistance coatings since they are tested using the same damage test setup and the same protocol. In 2009 a high reflector coating was selected at a wavelength of 786 nm at normal incidence at a pulse length of 180 femtoseconds. A double blind test assured sample and submitter anonymity so only a summary of the results are presented here. In addition to the laser resistance results, details of deposition processes, coating materials and layer count, and spectral results will also be shared.

  2. Damage spreading and coupling in Markov chains

    E-Print Network [OSTI]

    Etienne P. Bernard; Cédric Chanal; Werner Krauth

    2011-06-23

    In this paper, we relate the coupling of Markov chains, at the basis of perfect sampling methods, with damage spreading, which captures the chaotic nature of stochastic dynamics. For two-dimensional spin glasses and hard spheres we point out that the obstacle to the application of perfect-sampling schemes is posed by damage spreading rather than by the survey problem of the entire configuration space. We find dynamical damage-spreading transitions deeply inside the paramagnetic and liquid phases, and show that critical values of the transition temperatures and densities depend on the coupling scheme. We discuss our findings in the light of a classic proof that for arbitrary Monte Carlo algorithms damage spreading can be avoided through non-Markovian coupling schemes.

  3. Micropatterned cell arrays for detecting DNA damage

    E-Print Network [OSTI]

    Mittal, Sukant

    2008-01-01

    Numerous agents are capable of interacting with DNA and damaging it. Permanent changes in the DNA structure can be both mutagenic and cytotoxic; therefore, methods to measure the susceptibility of cells to mutations are ...

  4. Formation damage in underbalanced drilling operations 

    E-Print Network [OSTI]

    Reyes Serpa, Carlos Alberto

    2003-01-01

    Formation damage has long been recognized as a potential source of reduced productivity and injectivity in both horizontal and vertical wells. From the moment that the pay zone is being drilled until the well is put on production, a formation...

  5. Dealing with Storm-Damaged Trees 

    E-Print Network [OSTI]

    Kirk, Melanie; Taylor, Eric; Foster, C. Darwin

    2005-10-25

    -Damaged Trees Melanie R. Kirk, Extension Program Specialist, Eric L. Taylor, Assistant Professor and Extension Specialist, and C. Darwin Foster, Associate Department Head and Extension Program Leader for Forestry, The Texas A&M University System Downed trees...

  6. Controlled ion implant damage profile for etching

    DOE Patents [OSTI]

    Arnold, Jr., George W. (Tijeras, NM); Ashby, Carol I. H. (Edgewood, NM); Brannon, Paul J. (Albuquerque, NM)

    1990-01-01

    A process for etching a material such as LiNbO.sub.3 by implanting ions having a plurality of different kinetic energies in an area to be etched, and then contacting the ion implanted area with an etchant. The various energies of the ions are selected to produce implant damage substantially uniformly throughout the entire depth of the zone to be etched, thus tailoring the vertical profile of the damaged zone.

  7. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect (OSTI)

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  8. MS17AM LEAK CHECKER The MS17AM is a manual valve version of the MS17AB. Since it is a manual unit, some care must be

    E-Print Network [OSTI]

    Massey, Thomas N.

    MS17AM LEAK CHECKER The MS17AM is a manual valve version of the MS17AB. Since it is a manual unit, some care must be exercised in the sequencing of the valves. The most important things to remember are that during normal leak testing: 1. The THROTTLE VALVE is NEVER opened unless the ROUGH VALVE has been

  9. No damage to bulk storage but entire customer bases wiped out in storm

    SciTech Connect (OSTI)

    Not Available

    1992-10-01

    This paper reports that interviews with key LP-gas industry spokesmen in hurricane-ravaged South Florida following Andrew's terrifying visit presented a picture of unimaginable destruction and nearly immeasurable losses of property but miraculously, not a single significant incident involving leaks or the loss of bulk storage. In the worst reports of damage to propane company facilities, Homestead Gas loss at least one building and the Suburban/Petrolane office building in Homestead no longer exists. (Temporary office arrangements were established by Suburban/Petrolane.) The customer base of these companies has been hit very had. For a while, it appeared that one Suburban/Petrolane employee was unaccounted for but that report turned out to be false. Shortly after the storm, crews were out securing gas systems in whatever locations they could reach. In one instance in which regular travel proved impossible, the gas company was forced to travel the long way around-pulling resources out of its Key West district and going north to Homestead. It became necessary in many cases for personnel to purchase cellular phones in order to maintain contact between the office and field crew.

  10. Leak testing plan for the Oak Ridge National Laboratory liquid low-level waste systems (active tanks): Revision 2. Volume 1: Regulatory background and plan approach; Volume 2: Methods, protocols, and schedules; Volume 3: Evaluation of the ORNL/LT-823DP differential pressure leak detection method; Appendix to Revision 2: DOE/EPA/TDEC correspondence

    SciTech Connect (OSTI)

    Douglas, D.G.; Wise, R.F.; Starr, J.W.; Maresca, J.W. Jr. [Vista Research, Inc., Mountain View, CA (United States)

    1994-11-01

    This document, the Leak Testing Plan for the Oak Ridge National Laboratory Liquid Low-Level Waste System (Active Tanks), comprises three volumes. The first two volumes address the component-based leak testing plan for the liquid low-level waste system at Oak Ridge, while the third volume describes the performance evaluation of the leak detection method that will be used to test this system. Volume 1, describes that portion of the liquid low-level waste system at that will be tested; it provides the regulatory background, especially in terms of the requirements stipulated in the Federal Facilities Agreement, upon which the leak testing plan is based. Volume 1 also describes the foundation of the plan, portions of which were abstracted from existing federal documents that regulate the petroleum and hazardous chemicals industries. Finally, Volume 1 gives an overview the plan, describing the methods that will be used to test the four classes of components in the liquid low-level waste system. Volume 2 takes the general information on component classes and leak detection methods presented in Volume 1 and shows how it applies particularly to each of the individual components. A complete test plan for each of the components is presented, with emphasis placed on the methods designated for testing tanks. The protocol for testing tank systems is described, and general leak testing schedules are presented. Volume 3 describes the results of a performance evaluation completed for the leak testing method that will be used to test the small tanks at the facility (those less than 3,000 gal in capacity). Some of the details described in Volumes 1 and 2 are expected to change as additional information is obtained, as the viability of candidate release detection methods is proven in the Oak Ridge environment, and as the testing program evolves.

  11. Simulation of a small break loss of coolant accident conducted at the BETHSY Integral Test Facility 

    E-Print Network [OSTI]

    Bott, Charles Patrick

    1992-01-01

    systems where the interaction mechanisms themselves are complex and contain uncertainties. Early analysis of reactor accidents concentrated on severe accident scenarios in- volving large piping ruptures in the coolant system. These were seen... as the design basis accident for the systems, since, if the system could survive such a, scenario and maintain fuel cladding integrity, any smaller piping break would be a less severe ac- cident of one already evaluated. The duration of these large break loss...

  12. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect (OSTI)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  13. Temperature of aircraft cargo flame exposure during accidents involving fuel spills

    SciTech Connect (OSTI)

    Mansfield, J.A.

    1993-01-01

    This report describes an evaluation of flame exposure temperatures of weapons contained in alert (parked) bombers due to accidents that involve aircraft fuel fires. The evaluation includes two types of accident, collisions into an alert aircraft by an aircraft that is on landing or take-off, and engine start accidents. Both the B-1B and B-52 alert aircraft are included in the evaluation.

  14. Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel

    SciTech Connect (OSTI)

    Kass, Michael D; Theiss, Timothy J; Janke, Christopher James; Pawel, Steven J

    2012-07-01

    The Energy Independence and Security Act (EISA) of 2007 was enacted by Congress to move the nation toward increased energy independence by increasing the production of renewable fuels to meet its transportation energy needs. The law establishes a new renewable fuel standard (RFS) that requires the nation to use 36 billion gallons annually (2.3 million barrels per day) of renewable fuel in its vehicles by 2022. Ethanol is the most widely used renewable fuel in the US, and its production has grown dramatically over the past decade. According to EISA and RFS, ethanol (produced from corn as well as cellulosic feedstocks) will make up the vast majority of the new renewable fuel requirements. However, ethanol use limited to E10 and E85 (in the case of flex fuel vehicles or FFVs) will not meet this target. Even if all of the E0 gasoline dispensers in the country were converted to E10, such sales would represent only about 15 billion gallons per year. If 15% ethanol, rather than 10% were used, the potential would be up to 22 billion gallons. The vast majority of ethanol used in the United States is blended with gasoline to create E10, that is, gasoline with up to 10% ethanol. The remaining ethanol is sold in the form of E85, a gasoline blend with as much as 85% ethanol that can only be used in FFVs. Although DOE remains committed to expanding the E85 infrastructure, that market will not be able to absorb projected volumes of ethanol in the near term. Given this reality, DOE and others have begun assessing the viability of using intermediate ethanol blends as one way to transition to higher volumes of ethanol. In October of 2010, the EPA granted a partial waiver to the Clean Air Act allowing the use of fuel that contains up to 15% ethanol for the model year 2007 and newer light-duty motor vehicles. This waiver represents the first of a number of actions that are needed to move toward the commercialization of E15 gasoline blends. On January 2011, this waiver was expanded to include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a relatively short timeframe. Initially, these material studies included test fuels of Fuel C,

  15. Characterization of Vadose Zone Sediments from C Waste Management Area: Investigation of the C-152 Transfer Line Leak

    SciTech Connect (OSTI)

    Brown, Christopher F.; Serne, R. JEFFREY; Bjornstad, Bruce N.; Valenta, Michelle M.; Lanigan, David C.; Vickerman, Tanya S.; Clayton, Ray E.; Geiszler, Keith N.; Iovin, Cristian; Clayton, Eric T.; Kutynakov, I. V.; Baum, Steven R.; Lindberg, Michael J.; Orr, Robert D.

    2007-02-05

    A geologic/geochemical investigation in the vicinity of UPR-200-E-82 was performed using pairs of cone-penetrometer probe holes. A total of 41 direct-push cone-penetrometer borings (19 pairs to investigate different high moisture zones in the same sampling location and 3 individual) were advanced to characterize vadose zone moisture and the distribution of contaminants. A total of twenty sample sets, containing up to two split-spoon liners and one grab sample, were delivered to the laboratory for characterization and analysis. The samples were collected around the documented location of the C-152 pipeline leak, and created an approximately 120-ft diameter circle around the waste site. UPR-200-E-82 was a loss of approximately 2,600 gallons of Cs-137 Recovery Process feed solution containing an estimated 11,300 Ci of cesium-137 and 5 Ci of technetium-99. Several key parameters that are used to identify subsurface contamination were measured, including: water extract pH, electrical conductivity, nitrate, technetium-99, sodium, and uranium concentrations and technetium-99 and uranium concentrations in acid extracts. All of the parameters, with the exception of electrical conductivity, were elevated in at least some of the samples analyzed as part of this study. Specifically, soil pH was elevated (from 8.69 to 9.99) in five samples collected northeast and southwest of the C-152 pipeline leak. Similarly, samples collected from these same cone-pentrometer holes contained significantly more water-extractable sodium (more than 50 ?g/g of dry sediment), uranium (as much as 7.66E-01 ?g/g of dry sediment), nitrate (up to 30 ?g/g of dry sediment), and technetium-99 (up to 3.34 pCi/g of dry sediment). Most of the samples containing elevated concentrations of water-extractable sodium also had decreased levels of water extractable calcium and or magnesium, indicating that tank-related fluids that were high in sodium did seep into the vadose zone near these probe holes. Several of the samples containing high concentrations of water-leachable uranium also contained high pore water corrected alkalinity (3.26E+03 mg/L as CaCO3), indicating that the elevated water-leachable uranium could be an artifact of uranyl-carbonate complexation of naturally occurring labile uranium. However, a mass scan of the water extract containing the highest concentration of uranium was performed via inductively coupled mass spectrometry over the range of 230 to 240 atomic mass units, and a discernable peak was observed at mass 236. Although the data is considered qualitative, the presence of uranium-236 in the 1:1 sediment:water extract is a clear indication that the sample contains contaminant uranium [Hanford reprocessed fuel waste]. After evaluating all the characterization and analytical data, there is no question that the vadose zone surrounding the C-152 pipeline leak site has been contaminated by waste generally sent to tanks. The two zones or regions that contained the largest amount of contaminants, either in concentration or by occurrence of several key constituents/contaminants of concern, were located: 1) between the 241-C-151 and 241-C-152 Diversion Boxes (near the location of UPR-200-E-82) and 2) directly across the C-152 waste site near the C-153 Diversion Box (near where a pipeline, which connects the two diversion boxes, is shown on old blue prints . Without the use of more sophisticated analytical techniques, such as isotope signature analysis of ruthenium fission product isotopes, it is impossible to determine if the contamination observed at these two locations are from the same waste source or are a result of different leak events.

  16. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple...

    Energy Savers [EERE]

    Board concluded that the direct cause of the accident was the release of airborne contamination from a degraded package that contained cellulose material and plutonium-238...

  17. Integrating accident management issues in the design of future reactors EDF tentative approach

    SciTech Connect (OSTI)

    Berbey, P.; Vidard, M. [EDF-SEPTEN, Villeurbanne (France)

    1997-12-01

    In next generation plants, Severe Accidents will be explicitly considered at a very early stage of the design, and risk significant challenges will be addressed as appropriate. As design provisions could be considered to address some of these challenges, the need for Severe Accident Management (SAM) could be debated. For EDF, Accident Management (AM) in general, and SAM in particular, will remain a cornerstone of plant safety. Design provisions, if any, will have to be such as not to preclude any SAM measure with the potential for preventing accident sequences from progressing further. 3 refs., 2 figs.

  18. Level 1 Accident Report of the March 1, 2010 Bobcat Fatality...

    Energy Savers [EERE]

    Bluffs Substation March 31, 2010 On March 2, 2010 at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was...

  19. Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS

    Office of Energy Efficiency and Renewable Energy (EERE)

    DOE O 225.1B prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and...

  20. Summary of the SRS Severe Accident Analysis Program, 1987--1992

    SciTech Connect (OSTI)

    Long, T.A.; Hyder, M.L.; Britt, T.E.; Allison, D.K.; Chow, S.; Graves, R.D.; DeWald, A.B. Jr.; Monson, P.R. Jr.; Wooten, L.A.

    1992-11-01

    The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe accidents in SRS reactors that is adequate to support safety documentation for these reactors, including the Safety Analysis Report (SAR), the Probabilistic Risk Assessment (PRA), and other studies evaluating the safety of reactor operation; To provide tools and bases for the evaluation of existing or proposed safety related equipment in the SRS reactors; To provide bases for the development of accident management procedures for the SRS reactors; To develop and maintain on the site a sufficient body of knowledge, including documents, computer codes, and cognizant engineers and scientists, that can be used to authoritatively resolve questions or issues related to reactor accidents. The Severe Accident Analysis Program was instituted in 1987 and has already produced a substantial amount of information, and specialized calculational tools. Products of the Severe Accident Analysis Program (listed in Section 9 of this report) have been used in the development of the Safety Analysis Report (SAR) and the Probabilistic Risk Assessment (PRA), and in the development of technical specifications for the SRS reactors. A staff of about seven people is currently involved directly in the program and in providing input on severe accidents to other SRS activities.

  1. Double-Shell Tank Visual Inspection Changes Resulting from the Tank 241-AY-102 Primary Tank Leak

    SciTech Connect (OSTI)

    Girardot, Crystal L. [Washington River Protection Solutions, Richland, WA (United States); Washenfelder, Dennis J. [Washington River Protection Solutions, Richland, WA (United States); Johnson, Jeremy M. [USDOE Office of River Protection, Richland, WA (United States); Engeman, Jason K. [Washington River Protection Solutions, Richland, WA (United States)

    2013-11-14

    As part of the Double-Shell Tank (DST) Integrity Program, remote visual inspections are utilized to perform qualitative in-service inspections of the DSTs in order to provide a general overview of the condition of the tanks. During routine visual inspections of tank 241-AY-102 (AY-102) in August 2012, anomalies were identified on the annulus floor which resulted in further evaluations. In October 2012, Washington River Protection Solutions, LLC determined that the primary tank of AY-102 was leaking. Following identification of the tank AY-102 probable leak cause, evaluations considered the adequacy of the existing annulus inspection frequency with respect to the circumstances of the tank AY-102 1eak and the advancing age of the DST structures. The evaluations concluded that the interval between annulus inspections should be shortened for all DSTs, and each annulus inspection should cover > 95 percent of annulus floor area, and the portion of the primary tank (i.e., dome, sidewall, lower knuckle, and insulating refractory) that is visible from the annulus inspection risers. In March 2013, enhanced visual inspections were performed for the six oldest tanks: 241-AY-101, 241-AZ-101,241-AZ-102, 241-SY-101, 241-SY-102, and 241-SY-103, and no evidence of leakage from the primary tank were observed. Prior to October 2012, the approach for conducting visual examinations of DSTs was to perform a video examination of each tank's interior and annulus regions approximately every five years (not to exceed seven years between inspections). Also, the annulus inspection only covered about 42 percent of the annulus floor.

  2. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    SciTech Connect (OSTI)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  3. Analysis of Three Mile Island-Unit 2 accident

    SciTech Connect (OSTI)

    Not Available

    1980-03-01

    The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert.

  4. Advance plant severe accident/thermal hydraulic issues for ACRS

    SciTech Connect (OSTI)

    Kress, T.S.

    1994-09-01

    The ACRS has been reviewing various advance plant designs for certification. The most active reviews have been for the ABWR, AP600, and System 80+. We have completed the reviews for ABWR and System 80+ and are presently concentrating on AP600. The ACRS gave essentially unqualified certification approval for the two completed reviews, yet,,during the process of review a number of issues arose and the plant designs changed somewhat to accommodate some of the ACRS concerns. In this talk, I will describe some of the severe accident and thermal hydraulic related issues we discussed in our reviews.

  5. Locations of criticality alarms and nuclear accident dosimeters at Hanford

    SciTech Connect (OSTI)

    Not Available

    1992-08-01

    Hanford facilities that contain fissionable materials capable of achieving critical mass are monitored with nuclear accident dosimeters (NADS) in compliance with the requirements of DOE Order 5480.11, Chapter XI, Section 4.c. (DOE 1988). The US Department of Energy (DOE) Richland Field Office (RL) has assigned the responsibility for maintaining and evaluating the Hanford NAD system to the Instrumentation and External Dosimetry (I ED) Section of Pacific Northwest Laboratory's (PNL's) Health Physics Department. This manual provides a description of the Hanford NAD, criteria and instructions for proper NAD placement, and the locations of these dosimeters onsite.

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

    SciTech Connect (OSTI)

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

    1984-01-01

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

  7. Accident Investigation Report - Fire Report | 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (BillionProvedTravel TravelChallenges | Department of Energy ASHRAEUs About UstheAccident

  8. In a mining accident, first responders are working against

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverseIMPACT EVALUATION PLAN FOR THE SITE-218 58ImprovingIna mining accident, first

  9. Back-reactions, short-circuits, leaks and other energy wasteful reactions in biological electron transfer: Redox tuning to survive life in O2

    E-Print Network [OSTI]

    Review Back-reactions, short-circuits, leaks and other energy wasteful reactions in biological of pathways and the use of short circuits, back-reactions and side-paths, all of which compromise efficiency ancestors is explained as providing a protective back-reaction pathway. This ``sacrifice

  10. Prevention of Salt Damage inPrevention of Salt Damage in LimestoneLimestone

    E-Print Network [OSTI]

    Petta, Jason

    Prevention of Salt Damage inPrevention of Salt Damage in LimestoneLimestone Kathy Whitaker.jpg #12;Introduction: Sodium Sulfate Thenardite: Na2SO4 Mirabilite: Na2SO4·10H2O Salt exposure for 5 weeks the stone by capillary uptake of water containing the dissolved salt. Degradation of mortar. #12

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

    E-Print Network [OSTI]

    Gupta, Rakesh

    -structure predictive damage model and its integration into the development of a performance-based seismic design and refinement of modern strength-based seismic design codes were in use. One result of this significant damage-based seismic design PBSD philosophy for woodframe structures. Such phi- losophies are currently being developed

  12. DAMAGE DETECTION IN COMPOSITES BY NONCONTACT LASER Byeongjin Park

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    DAMAGE DETECTION IN COMPOSITES BY NONCONTACT LASER ULTRASONIC Byeongjin Park 1 , Hoon Sohn 1 author: p.malinowski@imp.gda.pl ABSTRACT This study proposes an instantaneous damage localization, are obtained. Then possible damage locations are estimated through time-of-flight triangulation of damage

  13. DAMAGE ESTIMATION USING MULTI-OBJECTIVE GENETIC ALGORITHMS Faisal Shabbir

    E-Print Network [OSTI]

    Boyer, Edmond

    DAMAGE ESTIMATION USING MULTI-OBJECTIVE GENETIC ALGORITHMS Faisal Shabbir 1 , Piotr Omenzetter 2 1.omenzetter@abdn.ac.uk ABSTRACT It is common to estimate structural damage severity by updating a structural model against experimental responses at different damage states. When experimental results from the healthy and damaged

  14. RUPTURE BY DAMAGE ACCUMULATION IN ROCKS David Amitrano

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    RUPTURE BY DAMAGE ACCUMULATION IN ROCKS David Amitrano LIRIGM, Université J. Fourier, Grenoble of rocks is associated with microcracks nucleation and propagation, i.e. damage. The accumulation of damage as strength and modulus. The damage process can be studied both statically by direct observation of thin

  15. Ductile damage parameters identification for cold metal forming applications

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Ductile damage parameters identification for cold metal forming applications Pierre damage mechanics is essential to predict failure during cold metal forming applications. Several damage models can be found in the literature. These damage models are coupled with the mechanical behavior so

  16. Structural damage detection using the holder exponent.

    SciTech Connect (OSTI)

    Farrar, C. R. (Charles R.); Do, N. B. (Nguyen B.); Green, S. R. (Scott R.); Schwartz, T. A. (Timothy A.)

    2002-01-01

    This paper implements a damage detection strategy that identifies damage sensitive features associated with nonlinearities. Some rion-linezlrities result from discontinuities introduced into the data by certain types of darnage. These discontinuities may also result from noise in the measured dynamic response data or can be caused by random excitation of the system. The Holder Exponent, which is a measure of the degree to which a signal is differentiable, is used to detect the discontinuities. By studying the Holder bponent as a function af time, a statistical model is developed that classifies changes in the Holder Exponent that are associated with clamage-induced discontinuities. The results show that for certain cases, the Holder Exponent is an effective technique to detect damage.

  17. Method to reduce damage to backing plate

    DOE Patents [OSTI]

    Perry, Michael D. (Livermore, CA); Banks, Paul S. (Livermore, CA); Stuart, Brent C. (Fremont, CA)

    2001-01-01

    The present invention is a method for penetrating a workpiece using an ultra-short pulse laser beam without causing damage to subsequent surfaces facing the laser. Several embodiments are shown which place holes in fuel injectors without damaging the back surface of the sack in which the fuel is ejected. In one embodiment, pulses from an ultra short pulse laser remove about 10 nm to 1000 nm of material per pulse. In one embodiment, a plasma source is attached to the fuel injector and initiated by common methods such as microwave energy. In another embodiment of the invention, the sack void is filled with a solid. In one other embodiment, a high viscosity liquid is placed within the sack. In general, high-viscosity liquids preferably used in this invention should have a high damage threshold and have a diffusing property.

  18. Thermal Damage Characterization of Energetic Materials

    SciTech Connect (OSTI)

    Hsu, P C; DeHaven, M R; Springer, H K; Maienschein, J L

    2009-08-14

    We conducted thermal damage experiments at 180?C on PBXN-9 and characterized its material properties. Volume expansion at high temperatures was very significant which led to a reduction in material density. 2.6% of weight loss was observed, which was higher than other HMX-based formulations. Porosity of PBXN-9 increased to 16% after thermal exposure. Small-scale safety tests (impact, friction, and spark) showed no significant sensitization when the damaged samples were tested at room temperature. Gas permeation measurements showed that gas permeability in damaged materials was several orders of magnitude higher than that in pristine materials. In-situ measurements of gas permeability and density were proved to be possible at higher temperatures.

  19. Seawater can damage Saudi sandstone oil reservoirs

    SciTech Connect (OSTI)

    Dahab, A.S. (King Saud Univ., Riyadh (SA))

    1990-12-10

    Experiments have shown that formation damage from waterflooding of the Aramco and Alkhafji sandstones of Saudi Arabia will not occur if the salinity of the injected brines is higher than 20% NaCl. Because the connate water in these reservoirs has a high salt content of up to 231,000 ppm, Saudi oil fields are almost always susceptible to formation damage when flooded with seawater (about 38,500 ppm). The productive behavior of a reservoir can be affected by clay crystals developed within rock pores.

  20. Resistance of cotton to pink bollworm damage 

    E-Print Network [OSTI]

    Brazzel, J. R.

    1956-01-01

    for resistance to pink bollworra damage, College Station, Texas, 1SAI . 12. Group comparisons analyses for a number of cottons with Deltapine TA based on number of larvae re? covered per gram of boll weight in greenhouse screening experiments for pink... screening experiment for resistance to pink bollworm damage, College Station, Texas, 195U- 1955: (A) Deltapine 15, (B) G^ thurberi, (C) Texas 389, (D) MW-298, (E) G^ stocksii, '(F) Stoneville 2-B x G. tomentosum, (G) Hexaploia Z-61j., (H) MW-1U7 . ? 5U...

  1. Gabriele Simi1 BaBar SVT: Radiation Damage and BaBar SVT: Radiation Damage and

    E-Print Network [OSTI]

    California at Santa Cruz, University of

    Gabriele Simi1 BaBar SVT: Radiation Damage and BaBar SVT: Radiation Damage and Other OperationalBar and SVTIntro to BaBar and SVT Radiation Environment Damage to Si Detectors Damage to Front End Electronics for Photon and KL detection DIRC for K/ separation DCH for charged particle tracking SVT for tracking

  2. To be submitted to Continuum Mechanics and Thermodynamics From the onset of damage to rupture: construction of responses with damage

    E-Print Network [OSTI]

    Boyer, Edmond

    To be submitted to Continuum Mechanics and Thermodynamics From the onset of damage to rupture: construction of responses with damage localization for a general class of gradient damage models Kim Pham solutions for the traction problem of an elastic damaging bar. This bar has a softening behavior which obeys

  3. Delay-active damage versus non-local enhancement for anisotropic damage dynamics computations with alternated loading

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Delay-active damage versus non-local enhancement for anisotropic damage dynamics computations, Laboratoire d'´etudes dynamiques F-91191 GIF-SUR-YVETTE Abstract Anisotropic damage thermodynamics framework of anisotropic visco-damage, by introducing a material strain rate effect in the cases of positive hydro- static

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

    SciTech Connect (OSTI)

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

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

  5. ATMOSPHERIC MODELING IN SUPPORT OF A ROADWAY ACCIDENT

    SciTech Connect (OSTI)

    Buckley, R.; Hunter, C.

    2010-10-21

    The United States Forest Service-Savannah River (USFS) routinely performs prescribed fires at the Savannah River Site (SRS), a Department of Energy (DOE) facility located in southwest South Carolina. This facility covers {approx}800 square kilometers and is mainly wooded except for scattered industrial areas containing facilities used in managing nuclear materials for national defense and waste processing. Prescribed fires of forest undergrowth are necessary to reduce the risk of inadvertent wild fires which have the potential to destroy large areas and threaten nuclear facility operations. This paper discusses meteorological observations and numerical model simulations from a period in early 2002 of an incident involving an early-morning multicar accident caused by poor visibility along a major roadway on the northern border of the SRS. At the time of the accident, it was not clear if the limited visibility was due solely to fog or whether smoke from a prescribed burn conducted the previous day just to the northwest of the crash site had contributed to the visibility. Through use of available meteorological information and detailed modeling, it was determined that the primary reason for the low visibility on this night was fog induced by meteorological conditions.

  6. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

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

    2013-09-01

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

  7. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    SciTech Connect (OSTI)

    Wenxi Tian; Suizheng Qiu; Guanghui Su; Dounan Jia [Xi'an Jiaotong University, 28 Xianning Road, Xi'an 710049 (China); Xingmin Liu - China Institute of Atomic Energy

    2006-07-01

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  8. Type B Accident Investigation Board Report on the August 5, 1998, Load Haul Dump Accident at U16b Tunnel, Nevada Test Site

    Broader source: Energy.gov [DOE]

    Thisis theType B Accident Investigation Board report of an industrial accident at the Nevada Test site (NTS), U16b tunnel in which a Bechtel Nevada (BN) employee suffered a compressed skull fracture as a result of being struck onthe head by a valve and fitting assembly on the end of a hose whichhad been broken from a water pipe by a moving piece of construction equipment.

  9. Potential Threats from a Likely Nuclear Power Plant Accident: a Climatological Trajectory Analysis

    E-Print Network [OSTI]

    Chen, Shu-Hua

    in the near future as insecure nuclear power plants with a high risk of accidents remain in the regionPotential Threats from a Likely Nuclear Power Plant Accident: a Climatological Trajectory Analysis at the Metsamor Nuclear Power Plant would influence all of Turkey. Furthermore, vulnerable regions in Turkey after

  10. Trans-oceanic transport of 137 Cs from the Fukushima nuclear accident

    E-Print Network [OSTI]

    Yu, Peter K.N.

    Trans-oceanic transport of 137 Cs from the Fukushima nuclear accident and impact of hypothetical Fukushima-like events of future nuclear plants in Southern China Ka-Ming Wai a,b, , Peter K.N. Yu b. · Observed and modeled 137 Cs concentrations were comparable for the Fukushima accident. · The maximum

  11. Emergency response to a highway accident in Springfield, Massachusetts, on December 16, 1991

    SciTech Connect (OSTI)

    Not Available

    1992-06-01

    On December 16, 1991, a truck carrying unirradiated (fresh) nuclear fuel was involved in an accident on US Interstate 91, in Springfield, Massachusetts. This report describes the emergency response measures undertaken by local, State, Federal, and private parties. The report also discusses ``lessons learned`` from the response to the accident and suggests areas where improvements might be made.

  12. Emergency response to a highway accident in Springfield, Massachusetts, on December 16, 1991

    SciTech Connect (OSTI)

    Not Available

    1992-06-01

    On December 16, 1991, a truck carrying unirradiated (fresh) nuclear fuel was involved in an accident on US Interstate 91, in Springfield, Massachusetts. This report describes the emergency response measures undertaken by local, State, Federal, and private parties. The report also discusses lessons learned'' from the response to the accident and suggests areas where improvements might be made.

  13. CNG buses fire safety: learnings from recent accidents in France and Germany

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    CNG buses fire safety: learnings from recent accidents in France and Germany Lionel PERRETTE Saarland Holding, Sulzbach Saar/ Germany ABSTRACT The use of CNG in bus and private vehicles is growing steadily. Recent fire accidents involving CNG buses have shown that tanks may explode though compliant

  14. Recovery sequences for a station blackout accident at the Grand Gulf Nuclear Station

    SciTech Connect (OSTI)

    Carbajo, J.J. [Martin Marietta Energy Systems, Oak Ridge, TN (United States)

    1995-12-31

    Recovery sequences for a low-pressure, short term, station blackout severe accident at the Grand Gulf power plant have been investigated using the computer code MELCOR, version 1.8.3 PN. This paper investigates the effect of reflood timing and mass flow rate on accident recovery.

  15. Undulator Radiation Damage Experience at LCLS

    SciTech Connect (OSTI)

    Nuhn, H. D.; Field, C.; Mao, S.; Levashov, Y.; Santana, M.; Welch, J. N.; Wolf, Z.

    2015-01-06

    The SLAC National Accelerator Laboratory has been running the Linac Coherent Light Source (LCLS), the first x-ray Free Electron Laser since 2009. Undulator magnet damage from radiation, produced by the electron beam traveling through the 133-m long straight vacuum tube, has been and is a concern. A damage measurement experiment has been performed in 2007 in order to obtain dose versus damage calibrations. Radiation reduction and detection devices have been integrated into the LCLS undulator system. The accumulated radiation dose rate was continuously monitored and recorded. In addition, undulator segments have been routinely removed from the beamline to be checked for magnetic (50 ppm, rms) and mechanic (about 0.25 µm, rms) changes. A reduction in strength of the undulator segments is being observed, at a level, which is now clearly above the noise. Recently, potential sources for the observed integrated radiation levels have been investigated. The paper discusses the results of these investigation as well as comparison between observed damage and measured dose accumulations and discusses, briefly, strategies for the new LCLS-II upgrade, which will be operating at more than 300 times larger beam rate.

  16. Nondestructive Damage Detection in General Beams 

    E-Print Network [OSTI]

    Dincal, Selcuk

    2010-12-08

    is to provide NDE methodologies that simultaneously identify the location, the extent, and the severity of damage in general beams. By general beams, we mean beyond Euler-Bernoulli beams (i.e. slender beams) to deep beams and stubby beams whose response may...

  17. Gas condensate damage in hydraulically fractured wells 

    E-Print Network [OSTI]

    Adeyeye, Adedeji Ayoola

    2004-09-30

    This project is a research into the effect of gas condensate damage in hydraulically fractured wells. It is the result of a problem encountered in producing a low permeability formation from a well in South Texas owned by the El Paso Production...

  18. How do energetic ions damage metallic surfaces?

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

    Osetskiy, Yury N.; Calder, Andrew F.; Stoller, Roger E.

    2015-02-20

    Surface modification under bombardment by energetic ions observed under different conditions in structural and functional materials and can be either unavoidable effect of the conditions or targeted modification to enhance materials properties. Understanding basic mechanisms is necessary for predicting properties changes. The mechanisms activated during ion irradiation are of atomic scale and atomic scale modeling is the most suitable tool to study these processes. In this paper we present results of an extensive simulation program aimed at developing an understanding of primary surface damage in iron by energetic particles. We simulated 25 keV self-ion bombardment of Fe thin films withmore »(100) and (110) surfaces at room temperature. A large number of simulations, ~400, were carried out allow a statistically significant treatment of the results. The particular mechanism of surface damage depends on how the destructive supersonic shock wave generated by the displacement cascade interacts with the free surface. Three basic scenarios were observed, with the limiting cases being damage created far below the surface with little or no impact on the surface itself, and extensive direct surface damage on the timescale of a few picoseconds. In some instances, formation of large vacancy loops beneath the free surface was observed, which may explain some earlier experimental observations.« less

  19. 1 GENERAL DESCRIPTION OF BAOTOU EARTHQUAKE DAMAGES

    E-Print Network [OSTI]

    Spencer Jr., Billie F.

    frame gallery (west) of No. 5 coke oven system of BISC. It acted in the violation of the national, water towers, bridges, substations, gas holders, open-hearth furnaces and reservoirs etc, 50% of them and heat supplying pipe networks, as well as gas pipelines for the municipal utilities. Damages

  20. An evaluation of operating speed reduction as a surrogate measure for accident experience on horizontal curves on two-lane rural highways 

    E-Print Network [OSTI]

    Anderson, Ingrid Bernice

    1993-01-01

    of accidents is relatively small. This problem is especially significant at rural locations where low traffic volumes require several years to establish a sufficient accident rate. A small sample of accident data may not accurately reveal the safety... to accident records. " One geometric measure which several studies have shown to have a relationship with accident experience is degree of curvature. Glennon (7) cites five studies which concluded that accident experience increased with increasing degree...