Science.gov

Sample records for accident core heatup

  1. TMI-2 accident: core heat-up analysis

    SciTech Connect

    Ardron, K.H.; Cain, D.G.

    1981-01-01

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

  2. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. )

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  3. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    SciTech Connect

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization.

  4. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    SciTech Connect

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.

  5. Heatup of the TMI-2 lower head during core relocation

    SciTech Connect

    Wang, S.K.; Sienicki, J.J.; Spencer, B.W.

    1989-01-01

    An analysis has been carried out to assess the potential of a melting attack upon the reactor vessel lower head and incore instrument nozzle penetration weldments during the TMI core relocation event at 224 minutes. Calculations were performed to determine the potential for molten corium to undergo breakup into droplets which freeze and form a debris bed versus impinging upon the lower head as one or more coherent streams. The effects of thermal-hydraulic interactions between corium streams and water inside the lower plenum, the effects of the core support assembly structure upon the corium, and the consequences of corium relocation by way of the core former region were examined. 19 refs., 24 figs.

  6. N Reactor core heatup sensitivity study for the 32-inch unit cell model

    SciTech Connect

    Martin, F.; Zimmerman, B.; Heard, F.

    1988-02-01

    A number of N Reactor core heatup studies have been performed using the TRUMP-BD computer code. These studies were performed to address questions concerning the dependency of results on potential variations in the material properties and/or modeling assumptions. This report described and documents a series of 31 TRUMP-BD runs that were performed to determine the sensitivity of calculated inner-fuel temperatures to a variety of TRUMP input parameters and also to a change in the node density in a high-temperature-gradient region. The results of this study are based on the 32-in. model. 18 refs., 17 figs., 2 tab.

  7. Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

    SciTech Connect

    Ball, S.J.; Nypaver, D.J.

    1994-06-01

    An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.

  8. An investigation of core liquid level depression in small break loss-of-coolant accidents

    SciTech Connect

    Schultz, R.R.; Watkins, J.C. ); Motley, F.E.; Stumpf, H. ); Chen, Y.S. . Div. of Systems Research)

    1991-08-01

    Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs.

  9. Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

    SciTech Connect

    White, M.D.; Lombardo, N.J.; Heard, F.J.; Ogden, D.M.; Quapp, W.J.

    1988-04-01

    Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and for uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.

  10. Assessment of CRBR core disruptive accident energetics

    SciTech Connect

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

    1984-03-01

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

  11. Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

    SciTech Connect

    Kroeger, P.G.; Chan, B.C.

    1985-01-01

    Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the CB atmosphere temperature and pressure, as overpressurization is the most likely failure mode.

  12. Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

    NASA Astrophysics Data System (ADS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

  13. Model for heat-up of structures in VICTORIA

    SciTech Connect

    Bixler, N.E.

    1993-12-01

    VICTORIA is a mechanistic computer code that treats fission product behavior in the reactor coolant system during a severe accident. During an accident, fission products that deposit on structural surfaces produce heat loads that can cause fission products to revaporize and possibly cause structures, such as a pipe, to fail. This mechanism had been lacking from the VICTORIA model. This report describes the structural heat-up model that has recently been implemented in the code. A sample problem shows that revaporization of fission products can occur as structures heat up due to radioactive decay. In the sample problem, the mass of deposited fission products reaches a maximum, then diminishes. Similarly, temperatures of the deposited film and adjoining structure reach a maximum, then diminish.

  14. Severe accident simulation at Olkiuoto

    SciTech Connect

    Tirkkonen, H.; Saarenpaeae, T.; Cliff Po, L.C.

    1995-09-01

    A personal computer-based simulator was developed for the Olkiluoto nuclear plant in Finland for training in severe accident management. The generic software PCTRAN was expanded to model the plant-specific features of the ABB Atom designed BWR including its containment over-pressure protection and filtered vent systems. Scenarios including core heat-up, hydrogen generation, core melt and vessel penetration were developed in this work. Radiation leakage paths and dose rate distribution are presented graphically for operator use in diagnosis and mitigation of accidents. Operating on an graphically for operator use in diagnosis and mitigation of accidents. Operating on an 486 DX2-66, PCTRAN-TVO achieves a speed about 15 times faster than real-time. A convenient and user-friendly graphic interface allows full interactive control. In this paper a review of the component models and verification runs are presented.

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

    SciTech Connect

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J.

    1994-02-01

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

  16. Safety evaluation of MHTGR licensing basis accident scenarios

    SciTech Connect

    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. Estimates of early containment loads from core melt accidents. Draft report for comment

    SciTech Connect

    1985-12-01

    The thermal-hydraulic processes and corium debris-material interactions that can result from core melting in a severe accident have been studied to evaluate the potential effect of such phenomena on containment integrity. Pressure and temperature loads associated with representative accident sequences have been estimated for the six various LWR containment types used within the United States. Summaries distilling the analyses are presented and an interpretation of the results provided. 13 refs., 68 figs., 39 tabs.

  18. Containment performance for the core melt accidents in BWRs with Mark I and Mark II containments

    SciTech Connect

    Perkins, K.R.; Yang, J.W.; Greene, G.A.; Pratt, W.T.; Hofmayer, C.

    1985-01-01

    Most previous risk assessment studies have assumed catastrophic failure of containments for severe accidents which are predicted to exceed the containment yield stress. This investigation analyzes the progression of a severe accident in order to develop realistic containment temperature and pressure loading, utilizes models for containment leakage estimates for the various loading histories, and assesses the expected failure modes and timing of releases for core melt accidents in Boiling Water Reactors (BWRs) with Mark I and Mark II containments. The results of the investigation indicate that leakage through the seal on the drywell head may be sufficient to prevent catastrophic failure of the containments for a wide range of hypothetical core melt scenarios. In addition, the investigation has indicated the potential for a previously inidentified failure mode (containment liner meltthrough) for Mark I containments in which a large fraction of the core is released from the vessel in a molten state. 14 refs.

  19. Teaching to the Common Core by Design, Not Accident

    ERIC Educational Resources Information Center

    Phillips, Vicki; Wong, Carina

    2012-01-01

    The Bill & Melinda Gates Foundation has created tools and supports intended to help teachers adapt to the Common Core State Standards in English language arts and mathematics. The tools seek to find the right balance between encouraging teachers' creativity and giving them enough guidance to ensure quality. They are the product of two years of…

  20. Incorporation of phenomenological uncertainties in probabilistic safety analysis - application to LMFBR core disruptive accident energetics

    SciTech Connect

    Najafi, B; Theofanous, T G; Rumble, E T; Atefi, B

    1984-08-01

    This report describes a method for quantifying frequency and consequence uncertainty distribution associated with core disruptive accidents (CDAs). The method was developed to estimate the frequency and magnitude of energy impacting the reactor vessel head of the Clinch River Breeder Plant (CRBRP) given the occurrence of hypothetical CDAs. The methodology is illustrated using the CRBR example.

  1. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

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

    SciTech Connect

    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.

  3. Development of integrated core disruptive accident analysis code for FBR - ASTERIA-FBR

    SciTech Connect

    Ishizu, T.; Endo, H.; Tatewaki, I.; Yamamoto, T.; Shirakawa, N.

    2012-07-01

    The evaluation of consequence at the severe accident is the most important as a safety licensing issue for the reactor core of liquid metal cooled fast breeder reactor (LMFBR), since the LMFBR core is not in an optimum condition from the viewpoint of reactivity. This characteristics might induce a super-prompt criticality due to the core geometry change during the core disruptive accident (CDA). The previous CDA analysis codes have been modeled in plural phases dependent on the mechanism driving a super-prompt criticality. Then, the following event is calculated by connecting different codes. This scheme, however, should introduce uncertainty and/or arbitrary to calculation results. To resolve the issues and obtain the consistent calculation results without arbitrary, JNES is developing the ASTERIA-FBR code for the purpose of providing the cross-check analysis code, which is another required scheme to confirm the validity of the evaluation results prepared by applicants, in the safety licensing procedure of the planned high performance core of Monju. ASTERIA-FBR consists of the three major calculation modules, CONCORD, dynamic-GMVP, and FEMAXI-FBR. CONCORD is a three-dimensional thermal-hydraulics calculation module with multi-phase, multi-component, and multi-velocity field model. Dynamic-GMVP is a space-time neutronics calculation module. FEMAXI-FBR calculates the fuel pellet deformation behavior and fuel pin failure behavior. This paper describes the needs of ASTERIA-FBR development, major module outlines, and the model validation status. (authors)

  4. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    SciTech Connect

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  5. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    SciTech Connect

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris.

  6. Large break loss-of-coolant accident analyses for the high flux isotope reactor

    SciTech Connect

    Taleyarkhan, R.P. )

    1989-01-01

    The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

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

    SciTech Connect

    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.

  8. On the effect of accident conditions on the molten core debris relocation into lower head of a PWR vessel

    NASA Astrophysics Data System (ADS)

    An, Xuegao

    From 1975 to present, it has been found that the primary risk to the public health and safety from nuclear power reactors lies in ``beyond design basis'' accidents. During such severe accidents, melting of the reactor core may lead to a loss of primary system integrity, or even containment failure, which will allow escape of significant amounts of radioactive material to the environment. It is very important to understand the mechanism of reactor core degradation during a severe accident. In this study, the damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out using the computer code SCDAP/RELAP5. Different modeling parameters or models were used in calculations by version MOD3.2. The cladding oxidation shell ``durability'' parameter, which can control the timing of fuel clad failure, was varied. The heat flux model of steady-state natural convection of the molten pool was changed. The ultimate strength of the crust supporting the molten pool was doubled. These changes were made to examine the effects on the calculated core damage, and the molten pool expansion and its slumping. Different accident scenarios were simulated. The HPI/makeup flow rates were changed. The timing of opening and closing the PORV was considered. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the PORV opening was also changed. The effects of these accident scenarios on accident progression and core damage process were studied. From the calculated results, it was concluded that the accurate modeling of core damage phenomena was very important to the prediction of the later stage of an accident. According to code MOD3.2, the molten material in a pool slumped to the lower head of the reactor vessel when the juncture of the top and side crusts failed after the

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

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D.; Dolan, B.W.; Minarick, J.W.

    1998-11-01

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

  10. TRACE/PARCS Core Modeling of a BWR/5 for Accident Analysis of ATWS Events

    SciTech Connect

    Cuadra A.; Baek J.; Cheng, L.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    The TRACE/PARCS computational package [1, 2] isdesigned to be applicable to the analysis of light water reactor operational transients and accidents where the coupling between the neutron kinetics (PARCS) and the thermal-hydraulics and thermal-mechanics (TRACE) is important. TRACE/PARCS has been assessed for itsapplicability to anticipated transients without scram(ATWS) [3]. The challenge, addressed in this study, is to develop a sufficiently rigorous input model that would be acceptable for use in ATWS analysis. Two types of ATWS events were of interest, a turbine trip and a closure of main steam isolation valves (MSIVs). In the first type, initiated by turbine trip, the concern is that the core will become unstable and large power oscillations will occur. In the second type,initiated by MSIV closure,, the concern is the amount of energy being placed into containment and the resulting emergency depressurization. Two separate TRACE/PARCS models of a BWR/5 were developed to analyze these ATWS events at MELLLA+ (maximum extended load line limit plus)operating conditions. One model [4] was used for analysis of ATWS events leading to instability (ATWS-I);the other [5] for ATWS events leading to emergency depressurization (ATWS-ED). Both models included a large portion of the nuclear steam supply system and controls, and a detailed core model, presented henceforth.

  11. Benchmarks of severe fuel damage tests using a stand-alone version of heatup from MAAP-DOE, Task 3. 6. 8

    SciTech Connect

    Sharon, A.; Burelbach, J.P.; Baumgarten, I.; Tashjian, V.S.; Hammersley, R.J.

    1990-03-01

    The results of three benchmark efforts are presented which assess the adequacy of the core HEATUP model that was developed for MAAP by the DOE in predicting the thermal-hydraulic response of a simulated LWR core under conditions leading to severe fuel damage. The model is applied to experiments performed at Idaho National Engineering Laboratory (PBF-SFD 1-1), Pacific Northwest Laboratory (FLHT-2), and Kernforschungszentrum Karlsruhe (CORA-12). Key quantities of interest in this study are the hydrogen generation rate, the bundle temperature distribution, and material relocation. Consistent good agreement is demonstrated between the model predictions and experimental data. 11 refs., 109 figs., 12 tabs.

  12. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    SciTech Connect

    Katoh, Yutai; Terrani, Kurt A.

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

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

    SciTech Connect

    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.

  14. Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction

    NASA Astrophysics Data System (ADS)

    Farmer, M. T.; Leibowitz, L.; Terrani, K. A.; Robb, K. R.

    2014-05-01

    Simple scoping models that can be used to evaluate ATF performance under severe accident conditions have been developed. The methodology provides a fundamental technical basis (a.k.a. metric) based on the thermodynamic boundary for evaluating performance relative to that of traditional Zr-based claddings. The initial focus in this study was on UO2 fuel with the advanced claddings 310 SS, D9, FeCrAl, and SiC. The evaluation considered only energy release with concurrent combustible gas production from fuel-cladding-coolant interactions and, separately, molten core-concrete interactions at high temperatures. Other important phenomenological effects that can influence the rate and extent of cladding decomposition (e.g., eutectic interactions, degradation of other core constituents) were not addressed. For the cladding types addressed, potential combustible gas production under both in-vessel and ex-vessel conditions was similar to that for Zr. However, exothermic energy release from cladding oxidation was substantially less for iron-based alloys (by at least a factor of 4), and modestly less (by ∼20%) for SiC. Data on SiC-clad UO2 fuel performance under severe accident conditions are sparse in the literature; thus, assumptions on the nature of the cladding decomposition process were made in order to perform this initial screening evaluation. Experimental data for this system under severe accident conditions is needed for a proper evaluation and comparison to iron-based claddings.

  15. Test Data for USEPR Severe Accident Code Validation

    SciTech Connect

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  16. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  17. Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition

    SciTech Connect

    Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

    2008-10-01

    The objective of this project was to perform stress analysis for graphite support structures of the General Atomics’ 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International –Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

  18. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    SciTech Connect

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.

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

    SciTech Connect

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W.

    1997-04-01

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

  20. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    SciTech Connect

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  1. Mechanical Testing of PMCs under Simulated Rapid Heat-Up Propulsion Environments. II; In-Plane Compressive Behavior

    NASA Technical Reports Server (NTRS)

    Stokes, Eric H.; Shin, E. Eugene; Sutter, James K.

    2003-01-01

    Carbon fiber thermoset polymer matrix composites (PMC) with high temperature polyimide based in-situ polymerized monomer reactant (PMR) resin has been used for some time in applications which can see temperatures up to 550 F. Currently, graphite fiber PMR based composites are used in several aircraft engine components including the outer bypass duct for the GE F-404, exit flaps for the P&W F-100-229, and the core cowl for the GE/Snecma CF6-80A3. Newer formulations, including PMR-II-50 are being investigated as potential weight reduction replacements of various metallic components in next generation high performance propulsion rocket engines that can see temperatures which exceed 550 F. Extensive FEM thermal modeling indicates that these components are exposed to rapid heat-up rates (up to -200 F/sec) and to a maximum temperature of around 600 F. Even though the predicted maximum part temperatures were within the capability of PW-II-50, the rapid heat-up causes significant through-thickness thermal gradients in the composite part and even more unstable states when combined with moisture. Designing composite parts for such extreme service environments will require accurate measurement of intrinsic and transient mechanical properties and the hygrothermal performance of these materials under more realistic use conditions. The mechanical properties of polymers degrade when exposed to elevated temperatures even in the absence of gaseous oxygen. Accurate mechanical characterization of the material is necessary in order to reduce system weight while providing sufficient factors of safety. Historically, the testing of PMCs at elevated temperatures has been plagued by the antagonism between two factors. First, moisture has been shown to profoundly affect the mechanical response of these materials at temperatures above their glass transition temperature while concurrently lowering the material's Tg. Moisture phenomena is due to one or a combination of three effects, i

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

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N.; Dolan, B.W.; Minarick, J.W. |

    1995-12-01

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

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

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.

    1997-04-01

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

  4. Analysis on the Density Driven Air-Ingress Accident in VHTRs

    SciTech Connect

    Eung Soo Kim; Chang Oh; Richard Schultz; David Petti

    2008-11-01

    Air-ingress following the pipe rupture is considered to be the most serious accident in the VHTRs due to its potential problems such as core heat-up, structural integrity and toxic gas release. Previously, it has been believed that the main air-ingress mechanism of this accident is the molecular diffusion process between the reactor core and the cavity. However, according to some recent studies, there is another fast air-ingress process that has not been considered before. It is called density-driven stratified flow. The potential for density-driven stratified air ingress into the VHTR following a large-break LOCA was first described in the NGNP Methods Technical Program based on stratified flow studies performed with liquid. Studies on densitygradient driven stratified flow in advanced reactor systems has been the subject of active research for well over a decade since density-gradient dominated stratified flow is an inherent characteristic of passive systems used in advanced reactors. Recently, Oh et al. performed a CFD analysis on the stratified flow in the VHTR, and showed that this effect can significantly accelerate the air-ingress process in the VHTRs. They also proposed to replace the original air-ingress scenario based on the molecular diffusion with the one based on the stratified flow. This paper is focusing on the effect of stratified flow on the results of the air-ingress accident in VHTR

  5. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    SciTech Connect

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  6. Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

    SciTech Connect

    Massara, S.; Schmitt, D.; Bretault, A.; Lemasson, D.; Darmet, G.; Verwaerde, D.; Struwe, D.; Pfrang, W.; Ponomarev, A.

    2012-07-01

    In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B{sub 4}C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed

  7. Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

    SciTech Connect

    Pfeiffer, P.A.; Kulak, R.F.

    1993-01-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

  8. Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

    SciTech Connect

    Pfeiffer, P.A.; Kulak, R.F.

    1993-06-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

  9. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N.; Kannan, S.E.

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  10. Materials Reliability Program: Development of a New Process for Calculating RPV Heat-Up and Cool-Down Curves - Proof of Concept

    SciTech Connect

    M. EricksonKirk

    2005-04-30

    A strategy and framework were developed for incorporating best-estimate, fracture toughness models and methodologies into procedures for fracture safety assessment of nuclear RPVs during normal heat-up and cool-down operations. The process included detailed process flow diagramming to identify all details of the current process for obtaining heat-up and cool-down curves.

  11. Interpretation of the results of the CORA-33 dry core BWR test

    SciTech Connect

    Ott, L.J.; Hagen, S.

    1993-11-01

    All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.

  12. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    SciTech Connect

    Myers, B.F.; Morrissey, R.E.

    1980-04-01

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

  13. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    SciTech Connect

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

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

    SciTech Connect

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N.; Dolan, B.W.; Jansen, J.M.; Minarick, J.W.; Lau, W.; Salyer, W.D.

    1993-12-01

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

  15. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  16. Candu 6 severe core damage accident consequence analysis for steam generator tube rupture scenario using MAAP4-CANDU V4.0.5A: preliminary results

    SciTech Connect

    Petoukhov, S.M.; Awadh, B.; Mathew, P.M.

    2006-07-01

    This paper describes the preliminary results of the consequence analysis for a generic AECL CANDU 6 station, when it undergoes a postulated, low probability Steam Generator multiple Tube Rupture (SGTR) severe accident with assumed unavailability of several critical plant safety systems. The Modular Accident Analysis Program for CANDU (MAAP4-CANDU) code was used for this analysis. The SGTR accident is assumed to begin with the guillotine rupture of 10 steam generator tubes in one steam generator in Primary Heat Transport System (PHTS) loop 1. For the reference case, the following systems were assumed unavailable: moderator and shield cooling, emergency core cooling, crash cool-down, and main and auxiliary feed water. Two additional cases were analyzed, one with the crash cool-down system available, and another with the crash cool-down and the auxiliary feed water systems available. The three scenarios considered in this study show that most of the initial fission product inventory would be retained within the containment by various fission product retention mechanisms. For the case where the crash cool-down system was credited but the auxiliary feed water systems were not credited, the total mass of volatile fission products released to the environment including stable and radioactive isotopes was about four times more than in the reference case, because fission products could be released directly from the PHTS to the environment through the Main Steam Safety Valves (MSSVs), bypassing the containment. For the case where the crash cool-down and auxiliary feed water systems were credited, the volatile fission product release to the environment was insignificant, because the fission product release was substantially mitigated by scrubbing in the water pool in the secondary side of the steam generator (SG). (authors)

  17. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    SciTech Connect

    Heard, F.J.

    1997-11-19

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity.

  18. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    SciTech Connect

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-07-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

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

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N.; Dolan, B.W.; Minarick, J.W. |

    1995-12-01

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

  20. Characterization of core debris/concrete interactions for the Advanced Neutron Source. ANS Severe Accident Analysis Program

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.

    1992-02-01

    This report provides the results of a recent study conducted to explore the molten core/concrete interaction (MCCI) issue for the Advanced Neutron Source (ANS). The need for such a study arises from the potential threats to reactor system integrity posed by MCCI. These threats include direct attack of the concrete basemat of the containment; generation and release of large quantities of gas that can pressurize the containment; the combustion threat of these gases; and the potential generation, release, and transport of radioactive aerosols to the environment.

  1. 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

    Chu, T.L.; Musicki, Z.; Kohut, P.

    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.

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

    SciTech Connect

    Chu, T.L.; Musicki, Z.; Kohut, P.

    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. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful.

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

    SciTech Connect

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

    1994-06-01

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

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

    SciTech Connect

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y.; Rashid, Y.R.

    1998-08-01

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

  5. Interpretation of experimental results from the CORA core melt progression experiments

    SciTech Connect

    Hohorst, J.K.; Allison, C.M.

    1991-01-01

    Data obtained from the CORA bundle heatup and melting experiments, performed at Kernforschungszentrum, Karlsruhe, Germany, are being analyzed at the Idaho National Engineering Laboratory. The analysis is being performed as part of a systematic review of core melt progression experiments for the United States Nuclear Regulatory Commission to (a) develop an improved understanding of important phenomena occurring during a severe accident, (b) to validate existing severe accident models, and (c) where necessary, develop improved models. An assessment of the variations in damage progression behavior because of variations in test parameters (a) bundle design and size, (b) system pressure, (c) slow cooling of the damaged bundles in argon versus rapid quenching in water, and (d) bundle inlet temperatures and flow rates is provided in the paper. The influence of uncertainties in important test conditions is also discussed. Specific results presented include (a) bundle temperature, (b) the onset and movement of the oxidation front within the bundle, (c) fuel rod ballooning and rod failure, and (d) melt relocation and associated material interactions between bundle components and structures. 12 refs., 16 figs., 2 tabs.

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

    SciTech Connect

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M.; Bley, D.; Johnson, D.

    1994-06-01

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

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

    SciTech Connect

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

    1994-06-01

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

  8. SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

    SciTech Connect

    Knudson, Darrell Lee; Ghan, Larry Scott; Dobbe, Charles Albin

    1998-09-01

    Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles, and the steam generator (SG) tubes. The potential for a steam generator tube rupture (SGTR) is of particular concern because fission products could be released to the environment through such a failure. The Nuclear Regulatory Commission (NRC) developed a program to address SG tube integrity issues in operating pressurized water reactors (PWRs) based on the possibility for environmental release. An extensive effort to evaluate the potential for accident-induced SGTRs using SCDAP/RELAP5 at the Idaho National Engineering and Environmental Laboratory (INEEL) was directed as one part of the NRC program. All SCDAP/RELAP5 calculations performed during the INEEL evaluation were based on station blackout accidents (and variations thereof) because those accidents are considered to be one of the more likely scenarios leading to natural circulation flows at temperatures and pressures that could threaten SG tube integrity (as well as the integrity of other vulnerable RCS pressure boundaries). Variations that were addressed included consideration of the effects of RCP seal leaks, intentional RCS depressurization through pressurizer PORVs, SG secondary depressurization, DC-HL bypass flows, U-tube SG sludge accumulation, and quenching of upper plenum stainless steel upon relocation to the lower head. Where available, experimental data was used to guide simulation of natural circulation flows. Independent reviews of the applicability of the natural circulation experimental data, the suitability of the code, and the adequacy of the modeling were completed and review recommendations were incorporated into the evaluation within budget and

  9. Core disruptive accident margin seal

    DOEpatents

    Golden, Martin P.

    1979-01-01

    Apparatus for sealing the annulus defined within a substantially cylindrical rotatable riser assembly and plug combination of a nuclear reactor closure head. The apparatus comprises an inflatable sealing mechanism disposed in one portion of the riser assembly near the annulus such that upon inflation the sealing mechanism is radially actuated against the other portion of the riser assembly thereby sealing the annulus. The apparatus further comprises a connecting mechanism which places one end of the sealing mechanism in fluid communication with the reactor cover gas so that overpressurization of the reactor cover gas will increase the radial actuation of the sealing mechanism thus enhancing sealing of the annulus.

  10. Core disruptive accident margin seal

    DOEpatents

    Garin, John; Belsick, James C.

    1978-01-01

    An apparatus for sealing the annulus defined between a substantially cylindrical rotatable first riser assembly and plug combination disposed in a substantially cylindrical second riser assembly and plug combination of a nuclear reactor system. The apparatus comprises a flexible member disposed between the first and second riser components and attached to a metal member which is attached to an actuating mechanism. When the actuating mechanism is not actuated, the flexible member does not contact the riser components thus allowing the free rotation of the riser components. When desired, the actuating mechanism causes the flexible member to contact the first and second riser components in a manner to block the annulus defined between the riser components, thereby sealing the annulus between the riser components.

  11. Core disruptive accident margin seal

    DOEpatents

    Garin, John

    1978-01-01

    An apparatus for sealing the annulus defined between a substantially cylindrical rotatable first riser assembly and plug combination disposed in a substantially cylindrical second riser assembly and plug combination of a nuclear reactor system. The apparatus comprises a flexible metal member having a first side attached to one of the riser components and a second side extending toward the other riser component and an actuating mechanism attached to the flexible metal member while extending to an accessible location. When the actuating mechanism is not activated, the flexible metal member does not contact the other riser component thus allowing the free rotation of the riser assembly and plug combination. When desired, the actuating mechanism causes the second side of the flexible metal member to contact the other riser component thereby sealing the annulus between the components.

  12. Characterization of the Transient Response of the ILS with One Module Installed to Heatup Changes in Power Level and Cooldown

    SciTech Connect

    K. G. Condie; C. M. Stoots; J. E. O'Brien; J. S. Herring

    2007-12-01

    This report provides documentation of the initial startup and testing of the first electrolysis module in the Idaho National Laboratory (INL) High Temperature Steam Electrolysis Integrated Laboratory Scale (ILS) facility. Initial shakedown testing of the INL ILS experimental facility commenced on August 22, 2007. This fulfilled a DOE Level 2 milestone. Heatup of the first ILS module started at approximately 4:10 PM on September 24, 2007. Initial module testing continued for 420 hours. The test average H2 production rate was approximately 1.3 Nm3/hr (0.116 kg H2/hr), with a peak measured value of over 2 Nm3/hr (0.179 kg H2/hr). Significant module performance degradation was observed over the first 250 hours, after which no further degradation was noted for the remainder of the test. Once all test objectives had been successfully met, the test was terminated in a controlled fashion. Discussion is included concerning several modifications that will be incorporated into the facility components to improve reliability and ease of operation for future long term testing.

  13. Thermal hydraulic features of the TMI accident

    NASA Astrophysics Data System (ADS)

    Tolman, B.

    1985-10-01

    The Three Mile island (TMI)-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident scenario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiments. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors.

  14. Accident investigation

    NASA Technical Reports Server (NTRS)

    Laynor, William G. Bud

    1987-01-01

    The National Transportation Safety Board (NTSB) has attributed wind shear as a cause or contributing factor in 15 accidents involving transport-categroy airplanes since 1970. Nine of these were nonfatal; but the other six accounted for 440 lives. Five of the fatal accidents and seven of the nonfatal accidents involved encounters with convective downbursts or microbursts. Of other accidents, two which were nonfatal were encounters with a frontal system shear, and one which was fatal was the result of a terrain induced wind shear. These accidents are discussed with reference to helping the aircraft to avoid the wind shear or if impossible to help the pilot to get through the wind shear.

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

    SciTech Connect

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

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

  16. 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

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

    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.

  17. HTTF Core Stress Analysis

    SciTech Connect

    Brian D. Hawkes; Richard Schultz

    2012-07-01

    In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

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

    SciTech Connect

    1993-12-01

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

  19. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

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

    SciTech Connect

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

    1995-09-01

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

  1. FY-09 Report: Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect

    Chang H. Oh; Eung S. Kim

    2009-12-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event following on the heels of a VHTR depressurization is a very important incident. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air will enter the core through the break, leading to oxidation of the in-core graphite structure and fuel. If this accident occurs, the oxidation will accelerate heat-up of the bottom reflector and the reactor core and will eventually cause the release of fission products. The potential collapse of the core bottom structures causing the release of CO and fission products is one of the concerns. Therefore, experimental validation with the analytical model and computational fluid dynamic (CFD) model developed in this study is very important. Estimating the proper safety margin will require experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. It will also require effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods Research and Development project. The second year of this three-year project (FY-08 to FY-10) was focused on (a) the analytical, CFD, and experimental study of air ingress caused by density-driven, stratified, countercurrent flow; (b) advanced graphite oxidation experiments and modeling; (c) experimental study of burn-off in the core bottom structures, (d) implementation of advanced

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

    SciTech Connect

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

    1995-12-31

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation.

  3. 1994 Accident sequence precursor program results

    SciTech Connect

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.

    1996-01-01

    The Accident Sequence Precursor (ASP) Program involves the systematic review and evaluation of operational events that have occurred at light-water reactors to identify and categorize precursors to potential severe core damage accident sequences. The results of the ASP Program are published in an annual report. The most recent report, which contains the analyses of the precursors for 1994, is NUREG/CR-4674, Vols. 21 and 22, Precursors to Potential Severe Core Damage Accidents: 1994, A Status Report, published in December 1995. This article provides an overview of the ASP review and evaluation process and a summary of the results for 1994. 12 refs., 2 figs., 4 tabs.

  4. 1995 Accident Sequence Precursor (ASP) Program results

    SciTech Connect

    Muhlheim, M.D.; Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; O`Reilly, P.D.; Dolan, B.W.; Minarick, J.W.

    1997-01-01

    The Accident Sequence Precursor (ASP) Program involves the systematic review and evaluation of operational events that have occurred at light-water reactors to identify and categorize precursors to potential severe core damage accident sequences. The results of the ASP Program are published in an annual report. The most recent report, which contains the precursors for 1995, is NUREG/CR-4674, Volume 23, Precursors to Potential Severe Core Damage Accidents: 1995, A Status Report, published in April 1997. This article provides an overview of the ASP review and evaluation process and a summary of the results for 1995.

  5. Catalog of PRA dominant accident sequence information

    SciTech Connect

    Cathey, N.G.; Krantz, E.A.; Poloski, J.P.; Shapiro, B.J.

    1985-07-01

    Information concerning the dominant accident sequences from twelve published probabilistic risk assessments (PRA) is cataloged in this report, which is published as a part of the Accident Sequence Evaluation Program (ASEP). The purpose of this report is to provide users of PRA information a single reference document. The cataloged results include plant operation information, core-melt frequency, event tree models, dominant factors affecting core-melt and sequence frequencies, and a description of each dominant accident sequence. The report provides a consistent set of insights on the factors that drive the dominant accident sequences. ASEP has reconstructed the PRA fault tree models at the system or train level of detail and requantified the sequence likelihoods to provide the consistent insights. This work provides the information for the other ASEP activities on accident likelihood assessment for the operating and near-term operating plants.

  6. BWR Core Heat Transfer Code System.

    Energy Science and Technology Software Center (ESTSC)

    1999-04-27

    Version 00 MOXY is used for the thermal analysis of a planar section of a boiling water reactor (BWR) fuel element during a loss-of-coolant accident (LOCA). The code emplyoys models that describe heat transfer by conduction, convection, and thermal radiation, and heat generation by metal-water reaction and fission product decay. Models are included for considering fuel-rod swelling and rupture, energy transport across the fuel-to-cladding gap, and the thermal response of the canister. MOXY requires thatmore » time-dependent data during the blowdown process for the power normalized to the steady-state power, for the heat-transfer coefficient, and for the fluid temperature be provided as input. Internal models provide these parameters during the heatup and emergency cooling phases.« less

  7. Radiation accidents.

    PubMed

    Saenger, E L

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity. PMID:3526994

  8. Radiation accidents

    SciTech Connect

    Saenger, E.L.

    1986-09-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity.

  9. The Chernobyl accident: Causes and consequences

    SciTech Connect

    Malinauskas, A.P.

    1987-01-01

    Two explosions, one immediately following the other, in Unit 4 of the Chernobyl nuclear power station in the Soviet Union signaled the worst disaster ever to befall the commercial nuclear power production industry. This accident, which occurred at 1:24 a.m. on April 26, 1986, resulted from an almost incredible series of operational errors associated, ironically, with an attempt to enhance the capability of the reactor to safely accommodate station blackout accidents (i.e., accidents arising from a loss of station electrical power). Disruption of the core, due to a prompt criticality excursion, resulted in the destruction of the core vault and reactor building and the sudden dispersal of about 3% of the fuel from the core region into the environment. Lesser but significant releases of radioactivity continued through May 6, 1986, before attempts to certain the radioactivity and cool the remnants of the core were successful. The amount and composition of material released in the course of the accident remain somewhat uncertain, and inconsistencies in the release estimates are evident. The Soviet estimates, in addition to the dispersal of about 3% of the fuel, include complete release of the noble gas core inventory, 20% of the fission product iodine inventory, 15% of the tellurium inventory, and 10 to 13% of the fission product cesium inventory. The iodine and cesium release estimates are not consistent with the noble gas values, and are as much as a factor of two less than some estimates made by experts outside the Soviet Union.

  10. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  11. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    SciTech Connect

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  12. World commercial aircraft accidents

    SciTech Connect

    Kimura, C.Y.

    1993-01-01

    This report is a compilation of all accidents world-wide involving aircraft in commercial service which resulted in the loss of the airframe or one or more fatality, or both. This information has been gathered in order to present a complete inventory of commercial aircraft accidents. Events involving military action, sabotage, terrorist bombings, hijackings, suicides, and industrial ground accidents are included within this list. Included are: accidents involving world commercial jet aircraft, world commercial turboprop aircraft, world commercial pistonprop aircraft with four or more engines and world commercial pistonprop aircraft with two or three engines from 1946 to 1992. Each accident is presented with information in the following categories: date of the accident, airline and its flight numbers, type of flight, type of aircraft, aircraft registration number, construction number/manufacturers serial number, aircraft damage, accident flight phase, accident location, number of fatalities, number of occupants, cause, remarks, or description (brief) of the accident, and finally references used. The sixth chapter presents a summary of the world commercial aircraft accidents by major aircraft class (e.g. jet, turboprop, and pistonprop) and by flight phase. The seventh chapter presents several special studies including a list of world commercial aircraft accidents for all aircraft types with 100 or more fatalities in order of decreasing number of fatalities, a list of collision accidents involving commercial aircrafts, and a list of world commercial aircraft accidents for all aircraft types involving military action, sabotage, terrorist bombings, and hijackings.

  13. Designing an Experimental "Accident"

    ERIC Educational Resources Information Center

    Picker, Lester

    1974-01-01

    Describes an experimental "accident" that resulted in much student learning, seeks help in the identification of nematodes, and suggests biology teachers introduce similar accidents into their teaching to stimulate student interest. (PEB)

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

    SciTech Connect

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. ); Medford, G.T. )

    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.

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Anshari, Rio

    2012-06-01

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

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

    SciTech Connect

    Kastenberg, W.E.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D.

    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.

  18. Visualization of Traffic Accidents

    NASA Technical Reports Server (NTRS)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

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

    SciTech Connect

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

    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.

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

    SciTech Connect

    Dukelow, J S; Harrison, D G; Morgenstern, M

    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.

  1. Accident tolerant fuels for LWRs: A perspective

    NASA Astrophysics Data System (ADS)

    Zinkle, S. J.; Terrani, K. A.; Gehin, J. C.; Ott, L. J.; Snead, L. L.

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  2. Laser accidents: Being Prepared

    SciTech Connect

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

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

    SciTech Connect

    Tusheva, P.; Schaefer, F.; Kliem, S.

    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)

  4. Accident mortality among children

    PubMed Central

    Swaroop, S.; Albrecht, R. M.; Grab, B.

    1956-01-01

    The authors present statistics on mortality from accidents, with special reference to those relating to the age-group 1-19 years. For a number of countries figures are given for the proportional mortality from accidents (the number of accident deaths expressed as a percentage of the number of deaths from all causes) and for the specific death-rates, per 100 000 population, from all causes of death, from selected causes, from all causes of accidents, and from various types of accident. From these figures it appears that, in most countries, accidents are becoming relatively increasingly prominent as a cause of death in childhood, primarily because of the conquest of other causes of death—such as infectious and parasitic diseases, which formerly took a heavy toll of children and adolescents—but also to some extent because the death-rate from motor-vehicle accidents is rising and cancelling out the reduction in the rate for other causes of accidental death. In the authors' opinion, further epidemiological investigations into accident causation are required for the purpose of devising quicker and more effective methods of accident prevention. PMID:13383361

  5. Airline accident response.

    PubMed

    Bettes, Thomas

    2002-01-01

    This article outlines government regulations affecting accident response and offers guidelines for airline contingency plans in the face of major air disasters, such as those encountered on September 11, 2001. The author also touches upon the role of the corporate medical department in accident investigation and victim identification. PMID:11872433

  6. Civil aircraft accident investigation.

    PubMed

    Haines, Daniel

    2013-01-01

    This talk reviews some historic aircraft accidents and some more recent. It reflects on the division of accident causes, considering mechanical failures and aircrew failures, and on aircrew training. Investigation results may lead to improved aircraft design, and to appropriate crew training. PMID:24057309

  7. Anatomy of an Accident.

    ERIC Educational Resources Information Center

    Mobley, Michael

    1984-01-01

    The findings of industrial safety engineers in the areas of accident causation and prevention are wholly applicable to adventure programs. Adventure education instructors can use safety engineering concepts to assess the risk in a particular activity, understand factors that cause accidents, and intervene to minimize injuries and damages if…

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

    SciTech Connect

    Kohut, P.

    1994-07-01

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

  9. Analysis of Credible Accidents for Argonaut Reactors

    SciTech Connect

    Hawley, S. C.; Kathern, R. L.; Robkin, M. A.

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.

  10. Time-dependent accident sequence analysis

    SciTech Connect

    Chu, T.L.

    1983-01-01

    One problem of the current event tree methodology is that the transitions between accident sequences are not modeled. The causes of transitions are mostly due to operator actions during an accident. A model for such transitions is presented. A generalized algorithm is used for quantification. In the more realistic accident analysis, the progression of the physical processes, which determines the time available for proper operators response, is modeled. Furthermore, the uncertainty associated with the physical modeling is considered. As an example, the approach is applied to analyze TMI-type accidents. Statistical evidence is collected and used in assessing the frequency of stuck-open pressure operated relief valve at B and W plants as well as the frequency of misdiagnosis. Statistical data are also used in modeling the timing of operator actions during the accident. A thermal code (CUT) is developed to determine the time at which the core uncovery occurs. A response surface is used to propagate the uncertainty associated with the thermal code.

  11. Persistence of airline accidents.

    PubMed

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. PMID:20618386

  12. Sleep related vehicle accidents.

    PubMed Central

    Horne, J. A.; Reyner, L. A.

    1995-01-01

    OBJECTIVES--To assess the incidence, time of day, and driver morbidity associated with vehicle accidents where the most likely cause was the driver falling asleep at the wheel. DESIGN--Two surveys were undertaken, in southwest England and the midlands, by using police databases or on the spot interviews. SUBJECTS--Drivers involved in 679 sleep related vehicle accidents. RESULTS--Of all vehicle accidents to which the police were summoned, sleep related vehicle accidents comprised 16% on major roads in southwest England, and over 20% on midland motorways. During the 24 hour period there were three major peaks: at around 0200, 0600, and 1600. About half these drivers were men under 30 years; few such accidents involved women. CONCLUSIONS--Sleep related vehicle accidents are largely dependent on the time of day and account for a considerable proportion of vehicle accidents, especially those on motorways and other monotonous roads. As there are no norms for the United Kingdom on road use by age and sex for time of day with which to compare these data, we cannot determine what the hourly exposure v risk factors are for these subgroups. The findings are in close agreement with those from other countries. PMID:7888930

  13. Loss of pumping accident limit calculation for Savannah River Reactor

    SciTech Connect

    Paul, P.K.; Barbour, K.L. )

    1992-01-01

    For the Savannah River Site production reactors, the design basis accident reactor power limit ensures that if a double-ended guillotine break (DEGB) in a secondary cooling water pipe were to occur, the reactor will shut down safely. The primary reactor coolant is heavy water (D{sub 2}O) with secondary light water (H{sub 2}O) cooling. The accident scenario is a DEGB in one of two secondary coolant inlet header pipes with several assumed single failures. The recycled primary coolant loses its cooling, and the reactor core temperature begins to rise. Another possible accident is a DEGB in one of two heat exchanger secondary coolant effluent header pipes. The inlet header break is slightly more limiting than the effluent header break. Upon break detection, emergency shutdown begins and the emergency cooling system (ECS) activates. The accident scenario was constructed with regard to physical, mechanical, and human factors. The computer code TRAC simulates the accident.

  14. Molten core retention assembly

    DOEpatents

    Lampe, Robert F.

    1976-06-22

    Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

  15. Safety Is No Accident.

    ERIC Educational Resources Information Center

    Christiansen, Monty L.

    1985-01-01

    Liability suits involving accidents in park and recreation areas are expensive and intangible costs are incalculable. Risk management practices related to park planning, personnel, and administrative practices are discussed. (MT)

  16. Accident resistant transport container

    DOEpatents

    Andersen, John A.; Cole, James K.

    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.

  17. Accident resistant transport container

    DOEpatents

    Anderson, J.A.; Cole, K.K.

    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.

  18. FATAL ACCIDENT REPORTING SYSTEM (FARS)

    EPA Science Inventory

    The Fatal Accident Reporting System (FARS) database consist of three relational tables, containing data on automobile accidents on public U.S. roads that resulted in the death of one or more people within 30 days of the accident. Truck and trailer accidents are also included.

  19. Accident management information needs

    SciTech Connect

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. )

    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.

  20. Core-melt source reduction system

    DOEpatents

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1995-04-25

    A core-melt source reduction system for ending the progression of a molten core during a core-melt accident and resulting in a stable solid cool matrix. The system includes alternating layers of a core debris absorbing material and a barrier material. The core debris absorbing material serves to react with and absorb the molten core such that containment overpressurization and/or failure does not occur. The barrier material slows the progression of the molten core debris through the system such that the molten core has sufficient time to react with the core absorbing material. The system includes a provision for cooling the glass/molten core mass after the reaction such that a stable solid cool matrix results. 4 figs.

  1. Core-melt source reduction system

    DOEpatents

    Forsberg, Charles W.; Beahm, Edward C.; Parker, George W.

    1995-01-01

    A core-melt source reduction system for ending the progression of a molten core during a core-melt accident and resulting in a stable solid cool matrix. The system includes alternating layers of a core debris absorbing material and a barrier material. The core debris absorbing material serves to react with and absorb the molten core such that containment overpressurization and/or failure does not occur. The barrier material slows the progression of the molten core debris through the system such that the molten core has sufficient time to react with the core absorbing material. The system includes a provision for cooling the glass/molten core mass after the reaction such that a stable solid cool matrix results.

  2. [Psychogenesis of accidents].

    PubMed

    Giannattasio, E; Nencini, R; Nicolosi, N

    1988-01-01

    After having carried out a historical review of industrial psychology with specific attention to the evolution of the concept of causality in accidents, the Authors formulate their work hypothesis from that research which take into highest consideration the executives' attitudes in the genesis of the accidents. As dogmatism appears to be one of the most negative of executives' attitudes, the Authors administered Rockeach's Scale to 130 intermediate executives from 6 industries in Latium and observed the frequency index for accidents and the morbidity index (absenteeism) of the 2149 workhand. The Authors assumed that to high degree of dogmatism on the executives' side should correspond o a higher level of accidents and absenteeism among the staff. The data processing revealed that, due to the type of machinery employed, three of the industries examined should be considered as High Risk Industrie (HRI), while the remaining three could be considered as Low Risk Industries (LRI): in fact, due to the different working conditions, a significant lower number of accidents occurred in last the three. A statistically significant correlation between the executives' dogmatism and the number of accidents among their workhand in the HRI has been noticed, while this has not been observed in the LRI. This confirms, as had already been pointed out by Gemelli in 1944, that some "objective conditions" are requested so that the accident may actually take place. On the other hand the morbidity index has not shown any difference related to the different kind of industries (HRI, LRI): in both cases statistically significant correlations were obtained between the executives' dogmatism and the staff's absenteeism. absenteeism.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:3154344

  3. Composite Cores

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.

  4. Severe Accident Test Station Activity Report

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  6. Injuries are not accidents

    PubMed Central

    Gutiérrez, María Isabel

    2014-01-01

    Injuries are the result of an acute exposure to exhort of energy or a consequence of a deficiency in a vital element that exceeds physiological thresholds resulting threatens life. They are classified as intentional or unintentional. Injuries are considered a global health issue because they cause more than 5 million deaths per year worldwide and they are an important contributor to the burden of disease, especially affecting people of low socioeconomic status in low- and middle-income countries. A common misconception exists where injuries are thought to be the same as accidents; however, accidents are largely used as chance events, without taken in consideration that all these are preventable. This review discusses injuries and accidents in the context of road traffic and emphasizes injuries as preventable events. An understanding of the essence of injuries enables the standardization of terminology in public use and facilitates the development of a culture of prevention among all of us. PMID:25386040

  7. Accidents in Childhood

    PubMed Central

    Keddy, J. Arthur

    1964-01-01

    The causes of injury to 17,141 children brought to the emergency department of a large pediatric hospital in one year were studied. The leading causes of injury were: falls, 5682; cuts or piercings, 1902; poisonings, 1597; and transportation accidents, 1368. Included in these are 587 falls on or down stairs, 401 cuts due to glass, 630 poisonings from household or workshop substances, 510 poisonings from salicylate tablets, and 449 accidents involving bicycles or tricycles. Other findings included 333 injuries to fingers or hands in doors, usually car doors; 122 instances of pulled arms; 384 ingestions and 53 inhalations of foreign bodies; 60 alleged sexual assaults, 58 chemical burns, 127 wringer injuries, and four attempted suicides. A rewarding opportunity in accident prevention exists for hospitals that undertake to compile and distribute pertinent source data. PMID:14201260

  8. The Fukushima radiation accident: consequences for radiation accident medical management.

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

    The March 2011 radiation accident in Fukushima, Japan, is a textbook example of a radiation accident of global significance. In view of the global dimensions of the accident, it is important to consider the lessons learned. In this context, emphasis must be placed on consequences for planning appropriate medical management for radiation accidents including, for example, estimates of necessary human and material resources. The specific characteristics of the radiation accident in Fukushima are thematically divided into five groups: the exceptional environmental influences on the Fukushima radiation accident, particular circumstances of the accident, differences in risk perception, changed psychosocial factors in the age of the Internet and globalization, and the ignorance of the effects of ionizing radiation both among the general public and health care professionals. Conclusions like the need for reviewing international communication, interfacing, and interface definitions will be drawn from the Fukushima radiation accident. PMID:22951483

  9. Physics in Accident Investigations.

    ERIC Educational Resources Information Center

    Brake, Mary L.

    1981-01-01

    Describes physics formulas which can be used by law enforcement officials to determine the possible velocity of vehicles involved in traffic accidents. These include, among others, the slide to stop-level road, slide to stop-sloping roadway, and slide to stop-two different surfaces formulas. (JN)

  10. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    SciTech Connect

    Siefken, Larry James

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

  11. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    SciTech Connect

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-03-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of {minus}3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept.

  12. Applying STAMP in Accident Analysis

    NASA Technical Reports Server (NTRS)

    Leveson, Nancy; Daouk, Mirna; Dulac, Nicolas; Marais, Karen

    2003-01-01

    Accident models play a critical role in accident investigation and analysis. Most traditional models are based on an underlying chain of events. These models, however, have serious limitations when used for complex, socio-technical systems. Previously, Leveson proposed a new accident model (STAMP) based on system theory. In STAMP, the basic concept is not an event but a constraint. This paper shows how STAMP can be applied to accident analysis using three different views or models of the accident process and proposes a notation for describing this process.

  13. [Equestrian accidents in children].

    PubMed

    Giebel, G; Braun, K; Mittelmeier, W

    1993-11-01

    In a retrospective study we reviewed 262 horse riding related injuries in children younger than 16 which were treated between 1975 and 1989 at the Section of Traumatology in the Department of Surgery, University Hospital Homburg/Saar. In 155 of these accidents, detailed information was gained via a questionnaire. The typical patient profile was that of young female equestrians with little experience and little weekly riding practice, without practicing falling-exercises and warming up often using different horses. At the time of the accident only 59% were wearing a head protection. Most accidents happened in the summer months in the afternoon during leisure riding on a large familiar horse in the riding hall. Apart from the typical accidents like falling of the horse (64.9%) and falling with the horse (5.7%) accidents in handling the horse were of special significance: Kick by horse's hoof (11.8%), being stepped by horse (3.8%), horsebite (7.3%) and injuries of horse's bridle had their own pattern of injuries. Injuries of the distal parts of the upper extremity are preeminent in falling of the horse, whilst in falling with the horse head injuries and shoulder injuries are preeminent. Remarkably often injuries of kick by horse's hoof were causing sometimes even dangerous head injuries (41.6%). Overall in horse riding related injuries in childhood superficial soft tissue injuries (48.6%) and fractures (30.6%) were predominant. Fractures of the clavicle which are well known as a riding injury proved to be typical for a fall with the horse, whilst a fractured vertebra was only seen once amongst the 262 children treated. The severity of the injuries was lower than expected: In 85.1% of all the injuries only one body region was injured, 90.1% could be assigned to an injury severity score (ISS) of 1-3. Ponyriders had less severe injuries than riders of large horses. One fatal accident happened in handling a horse, in these situations preventive measures are often

  14. 49 CFR 835.11 - Obtaining Board accident reports, factual accident reports, and supporting information.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Obtaining Board accident reports, factual accident... Board accident reports, factual accident reports, and supporting information. It is the responsibility... obtain Board accident reports, factual accident reports, and accompanying accident docket files....

  15. 49 CFR 835.11 - Obtaining Board accident reports, factual accident reports, and supporting information.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Obtaining Board accident reports, factual accident... Board accident reports, factual accident reports, and supporting information. It is the responsibility... obtain Board accident reports, factual accident reports, and accompanying accident docket files....

  16. Accident Flying Squad

    PubMed Central

    Snook, Roger

    1972-01-01

    This paper describes the organization, evaluation, and costing of an independently financed and operated accident flying squad. 132 accidents involving 302 casualties were attended, six deaths were prevented, medical treatment contributed to the survival of a further four, and the condition or comfort of many other casualties was improved. The calls in which survival was influenced were evenly distributed throughout the three-and-a-half-year survey and seven of the 10 so aided were over 16 and under 30 years of age, all 10 being in the working age group. The time taken to provide the service was not excessive and the expense when compared with the overall saving was very small. The scheme was seen to be equally suitable for basing on hospital or general practice or both, and working as an integrated team with the ambulance service. The use of specialized transport was found to be unnecessary. Other benefits of the scheme included use of the experience of attending accidents to ensure relevant and realistic training for emergency service personnel, and an appreciation of the effect of ambulance design on the patient. ImagesFIG. 1FIG. 4 PMID:5069642

  17. ORNL analyses of AVR performance and safety

    SciTech Connect

    Cleveland, J.C.

    1985-01-01

    Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a hypothetical depressurized core heatup accident and consideration of how a depressurized core heatup test might be conducted by AVR staff. Also presented are initial analyses of a test involving a reduction in core flow and of a test involving reactivity insertion via control rod withdrawal.

  18. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect

    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.

  19. [Chernobyl nuclear power plant accident and Tokaimura criticality accident].

    PubMed

    Takada, Jun

    2012-03-01

    It is clear from inspection of historical incidents that the scale of disasters in a nuclear power plant accident is quite low level overwhelmingly compared with a nuclear explosion in nuclear war. Two cities of Hiroshima and Nagasaki were destroyed by nuclear blast with about 20 kt TNT equivalent and then approximately 100,000 people have died respectively. On the other hand, the number of acute death is 30 in the Chernobyl nuclear reactor accident. In this chapter, we review health hazards and doses in two historical nuclear incidents of Chernobyl and Tokaimura criticality accident and then understand the feature of the radiation accident in peaceful utilization of nuclear power. PMID:22514916

  20. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect

    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

  1. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    SciTech Connect

    Robb, Kevin R.

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, Fe

  2. DUBLIN CORE

    EPA Science Inventory

    The Dublin Core is a metadata element set intended to facilitate discovery of electronic resources. It was originally conceived for author-generated descriptions of Web resources, and the Dublin Core has attracted broad ranging international and interdisciplinary support. The cha...

  3. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  4. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    SciTech Connect

    Siefken, L.J.

    1998-08-01

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert`s law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions.

  5. Rear-end accident victims. Importance of understanding the accident.

    PubMed Central

    Sehmer, J. M.

    1993-01-01

    Family physicians regularly treat victims of rear-end vehicle accidents. This article describes how taking a detailed history of the accident and understanding the significance of the physical events is helpful in understanding and anticipating patients' morbidity and clinical course. Eight questions to ask patients are suggested to help physicians understand the severity of injury. PMID:8495140

  6. Reactor Accident Consequence Code

    SciTech Connect

    2015-11-02

    MACCS1.5 performs probabilistic calculations of potential off site consequences of the atmospheric releases of radioactive material in reactor accidents. The principal phenomena considered in MACCS are atmospheric transport, environmental contamination, emergency response, long term mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. MACCS can be used for a variety of applications including probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost benefit analysis. The time scale after the accident is divided into three phases: emergency, intermediate, and long term. The region surrounding the reactor is divided into a polar-coordinate grid, with the reactor located at the center, for the calculations. Two preprocessors, MAXGC and DOSFAC, are included. MAXGC generates the maximum allowable ground concentrations based on protective action guide (PAG) dose levels. DOSFAC generates the dose conversion data used by MACCS.

  7. Reactor Accident Consequence Code

    Energy Science and Technology Software Center (ESTSC)

    2015-11-02

    MACCS1.5 performs probabilistic calculations of potential off site consequences of the atmospheric releases of radioactive material in reactor accidents. The principal phenomena considered in MACCS are atmospheric transport, environmental contamination, emergency response, long term mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. MACCS can be used for a variety of applications including probabilistic risk assessment (PRA) ofmore » nuclear power plants and other nuclear facilities, sensitivity studies to gain a better understanding of the parameters important to PRA, and cost benefit analysis. The time scale after the accident is divided into three phases: emergency, intermediate, and long term. The region surrounding the reactor is divided into a polar-coordinate grid, with the reactor located at the center, for the calculations. Two preprocessors, MAXGC and DOSFAC, are included. MAXGC generates the maximum allowable ground concentrations based on protective action guide (PAG) dose levels. DOSFAC generates the dose conversion data used by MACCS.« less

  8. Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect

    Chang Ho Oh; Eung Soo Kim; Hee Cheon No; Nam Zin Cho

    2008-12-01

    The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) are very high priority for the NGNP program. Following a loss of coolant and system depressurization, air will enter the core through the break. Air ingress leads to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heat-up of the bottom reflector and the reactor core and will cause the release of fission products eventually. The potential collapse of the bottom reflector because of burn-off and the release of CO lead to serious safety problems. For estimation of the proper safety margin we need experimental data and tools, including accurate multi-dimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. We also need to develop effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods R&D project. This project is focused on (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the bottom reflector, (d) structural tests of the burnt-off bottom reflector, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i

  9. MNSR transient analyses and thermal-hydraulic safety margins for HEU and LEU cores using PARET

    SciTech Connect

    Olson, Arne P.; Jonah, S.A.

    2008-07-15

    Thermal-hydraulic performance characteristics of Miniature Neutron Source Reactors under long-term steady-state and transient conditions are investigated. Safety margins and limiting conditions attained during these events are determined. Modeling extensions are presented that enable the PARET/ANL code to realistically track primary loop heatup, heat exchange to the pool, and heat loss from the pool to air over the pool. Comparisons are made of temperature predictions for HEU and LEU fueled cores under transient conditions. Results are obtained using three different natural convection heat transfer correlations: the original (PARET/ANL version 5), Churchill-Chu, and an experiment- based correlation from the China Institute of Atomic Energy (CIAE). The MNSR, either fueled by HEU or by LEU, satisfies the design limits for long-term transient operation. (author)

  10. German aircraft accident statistics, 1930

    NASA Technical Reports Server (NTRS)

    Weitzmann, Ludwig

    1932-01-01

    The investigation of all serious accidents, involving technical defects in the airplane or engine, is undertaken by the D.V.L. in conjunction with the imperial traffic minister and other interested parties. All accidents not clearly explained in the reports are subsequently cleared up.

  11. Weather types and traffic accidents.

    PubMed

    Klaić, Z B

    2001-06-01

    Traffic accident data for the Zagreb area for the 1981-1982 period were analyzed to investigate possible relationships between the daily number of accidents and the weather conditions that occurred for the 5 consecutive days, starting two days before the particular day. In the statistical analysis of low accident days weather type classification developed by Poje was used. For the high accident days a detailed analyses of surface and radiosonde data were performed in order to identify possible front passages. A test for independence by contingency table confirmed that conditional probability of the day with small number of accidents is the highest, provided that one day after it "N" or "NW" weather types occur, while it is the smallest for "N1" and "Bc" types. For the remaining 4 days of the examined periods dependence was not statistically confirmed. However, northern ("N", "NE" and "NW") and anticyclonic ("Vc", "V4", "V3", "V2" and "mv") weather types predominated during 5-days intervals related to the days with small number of accidents. On the contrary, the weather types with cyclonic characteristics ("N1", "N2", "N3", "Bc", "Dol1" and "Dol"), that are generally accompanied by fronts, were the rarest. For 85% days with large number of accidents, which had not been caused by objective circumstances (such as poor visibility, damaged or slippery road etc.), at least one front passage was recorded during the 3-days period, starting one day before the day with large number of accidents. PMID:11787547

  12. First Responders and Criticality Accidents

    SciTech Connect

    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.

  13. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  14. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  15. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...) Explosion or fire not intentionally set by the operator. (b) Release of 5 gallons (19 liters) or more...

  16. Thermal-hydraulic studies on molten core-concrete interactions

    SciTech Connect

    Greene, G.A.

    1986-10-01

    This report discusses studies carried out in connection with light water power reactor accidents. Recent assessments have indicated that the consequences of molten-core concrete interactions dominate the considerations of severe accidents. The two areas of interest that have been investigated are interlayer heat and mass transfer and liquid-liquid boiling. Interlayer heat and mass transfer refers to processes that occur within a core melt between the stratified, immiscible phases of core oxides and metals. Liquid-liquid boiling refers to processes that occur at the melt-concrete on melt-coolant interface. (JDH)

  17. 24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  18. Industrial accidents triggered by lightning.

    PubMed

    Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

    2010-12-15

    Natural disasters can cause major accidents in chemical facilities where they can lead to the release of hazardous materials which in turn can result in fires, explosions or toxic dispersion. Lightning strikes are the most frequent cause of major accidents triggered by natural events. In order to contribute towards the development of a quantitative approach for assessing lightning risk at industrial facilities, lightning-triggered accident case histories were retrieved from the major industrial accident databases and analysed to extract information on types of vulnerable equipment, failure dynamics and damage states, as well as on the final consequences of the event. The most vulnerable category of equipment is storage tanks. Lightning damage is incurred by immediate ignition, electrical and electronic systems failure or structural damage with subsequent release. Toxic releases and tank fires tend to be the most common scenarios associated with lightning strikes. Oil, diesel and gasoline are the substances most frequently released during lightning-triggered Natech accidents. PMID:20817399

  19. Columbia Accident Investigation Board

    NASA Technical Reports Server (NTRS)

    2003-01-01

    The Columbia Accident Investigation Board gathers for a second day for its third public hearing, held in Cape Canaveral, Florida. The CAIB was set up to examine STS-107 and analyze exploratory tests. Navy Admiral Harold W. 'Hal' Gehman Jr. was designated as the Chairman of the Board. From left to right in this photo sit Board Members Steven B. Wallace, Scott Hubbard, Dr. John Logsdon, Rear Admiral Stephen Turcotte, Hal Gehman, General Duane Deal, Dr. Douglas Osheroff, and Maj. General Kenneth W. Hess. Not shown are Maj. General John Barry, Dr. James N. Hallock, Roger Tetrault, Dr. Sheila Widnall, and Dr. Sally Ride. For more information on STS-107, please see GRIN Columbia General Explanation

  20. [The radiation accident].

    PubMed

    Stögmann, W

    1988-08-26

    The reactor accident of Chernobyl in April 1986 has shown us all the dangers which are inherent ever in the peaceful use of atomic energy. The effects of exposure to ionizing radiation are dependent on biological effectiveness, on dose, on duration of exposure and on the age of the exposed person (the younger the graver). Acute ionizing radiation of the whole body leads to radiation disease or radiation syndrome of different stages of severity according to dosage. If the patient survives other consequences of ionizing radiation may arise: non-stochastic effects such as cataracts, keloid formation, fibrosis of the lungs and infertility) and stochastic effects (oncogenesis and mutagenesis). The sensitivity to ionizing radiation is especially high in childhood because of the high velocity of cell metabolism and cell growth, the large body-surface area and because their repair mechanism following radiation damage is not yet. PMID:3188527

  1. Transport aircraft accident dynamics

    NASA Technical Reports Server (NTRS)

    Cominsky, A.

    1982-01-01

    A study was carried out of 112 impact survivable jet transport aircraft accidents (world wide) of 27,700 kg (60,000 lb.) aircraft and up extending over the last 20 years. This study centered on the effect of impact and the follow-on events on aircraft structures and was confined to the approach, landing and takeoff segments of the flight. The significant characteristics, frequency of occurrence and the effect on the occupants of the above data base were studied and categorized with a view to establishing typical impact scenarios for use as a basis of verifying the effectiveness of potential safety concepts. Studies were also carried out of related subjects such as: (1) assessment of advanced materials; (2) human tolerance to impact; (3) merit functions for safety concepts; and (4) impact analysis and test methods.

  2. BNL severe-accident sequence experiments and analysis program. [PWR; BWR

    SciTech Connect

    Greene, G.A.; Ginsberg, T.; Tutu, N.K.

    1983-01-01

    In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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)

  5. [Hidden statistics of traffic accidents].

    PubMed

    Nordentoft, E L; Larsen, C F; Jørgensen, H R

    1989-10-23

    Only 19% of the 3,071 injured persons who were treated in the casualty department of Odense Hospital following traffic accidents in 1987 could be found again in the police registers of traffic accidents from the same region. All of the registrations from the police registers from the central region could be found again in the casualty department. In 1971, the corresponding coverage was 36%. The degree of coverage is particularly low for single bicycle accidents, other bicycle accidents, other single accidents and the hours immediately after midnight. Considerable disagreement exists concerning registration of the use of safety belts and crash helmets. In Odense, the municipal road authorities utilize the localization of the accidents reported by the casualty department. The decrease in the degree of coverage is due mainly to an increasing proportion of bicycle accidents. Where casualties require admission to hospital, the coverage is approximately 75%. This has remained unchanged throughout the years and it is therefore suggested that this proportion should be employed as indicator of the effect of the majority of prophylactic measures. In addition, proposals are made for simplification of the police registration forms. PMID:2588362

  6. Core strengthening.

    PubMed

    Arendt, Elizabeth A

    2007-01-01

    Several recent studies have evaluated interventional techniques designed to reduce the risk of serious knee injuries, particularly noncontact anterior cruciate ligament injuries in female athletes. Maintenance of rotational control of the limb underneath the pelvis, especially in response to cutting and jumping activities, is a common goal in many training programs. Rotational control of the limb underneath the pelvis is mediated by a complex set of factors including the strength of the trunk muscles and the relationship between the core muscles. It is important to examine the interrelationship between lower extremity function and core stability. PMID:17472321

  7. A review of criticality accidents

    SciTech Connect

    Stratton, W R; Smith, D R

    1989-03-01

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

  8. Accident sequence precursor analysis level 2/3 model development

    SciTech Connect

    Lui, C.H.; Galyean, W.J.; Brownson, D.A.

    1997-02-01

    The US Nuclear Regulatory Commission`s Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models.

  9. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Kunkel, D.; Lelieveld, J.; Lawrence, M. G.

    2012-04-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90 % of emitted 137Cs would be transported beyond 50 km and about 50 % beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  10. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2011-11-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50km and about 50% beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  11. Advanced accident sequence precursor analysis level 2 models

    SciTech Connect

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L.

    1996-03-01

    The U.S. Nuclear Regulatory Commission Accident Sequence Precursor program pursues the ultimate objective of performing risk significant evaluations on operational events (precursors) occurring in commercial nuclear power plants. To achieve this objective, the Office of Nuclear Regulatory Research is supporting the development of simple probabilistic risk assessment models for all commercial nuclear power plants (NPP) in the U.S. Presently, only simple Level 1 plant models have been developed which estimate core damage frequencies. In order to provide a true risk perspective, the consequences associated with postulated core damage accidents also need to be considered. With the objective of performing risk evaluations in an integrated and consistent manner, a linked event tree approach which propagates the front end results to back end was developed. This approach utilizes simple plant models that analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude and timing of a radioactive release to the environment, and calculate the consequences for a given release. Detailed models and results from previous studies, such as the NUREG-1150 study, are used to quantify these simple models. These simple models are then linked to the existing Level 1 models, and are evaluated using the SAPHIRE code. To demonstrate the approach, prototypic models have been developed for a boiling water reactor, Peach Bottom, and a pressurized water reactor, Zion.

  12. [Prevention of bicycle accidents].

    PubMed

    Zwipp, H; Barthel, P; Bönninger, J; Bürkle, H; Hagemeister, C; Hannawald, L; Huhn, R; Kühn, M; Liers, H; Maier, R; Otte, D; Prokop, G; Seeck, A; Sturm, J; Unger, T

    2015-04-01

    For a very precise analysis of all injured bicyclists in Germany it would be important to have definitions for "severely injured", "seriously injured" and "critically injured". By this, e.g., two-thirds of surgically treated bicyclists who are not registered by the police could become available for a general analysis. Elderly bicyclists (> 60 years) are a minority (10 %) but represent a majority (50 %) of all fatalities. They profit most by wearing a helmet and would be less injured by using special bicycle bags, switching on their hearing aids and following all traffic rules. E-bikes are used more and more (145 % more in 2012 vs. 2011) with 600,000 at the end of 2011 and are increasingly involved in accidents but still have a lack of legislation. So even for pedelecs 45 with 500 W and a possible speed of 45 km/h there is still no legislative demand for the use of a protecting helmet. 96 % of all injured cyclists in Germany had more than 0.5 ‰ alcohol in their blood, 86 % more than 1.1 ‰ and 59 % more than 1.7 ‰. Fatalities are seen in 24.2 % of cases without any collision partner. Therefore the ADFC calls for a limit of 1.1 ‰. Some virtual studies conclude that integrated sensors in bicycle helmets which would interact with sensors in cars could prevent collisions or reduce the severity of injury by stopping the cars automatically. Integrated sensors in cars with opening angles of 180° enable about 93 % of all bicyclists to be detected leading to a high rate of injury avoidance and/or mitigation. Hanging lamps reduce with 35 % significantly bicycle accidents for children, traffic education for children and special trainings for elderly bicyclists are also recommended as prevention tools. As long as helmet use for bicyclists in Germany rates only 9 % on average and legislative orders for using a helmet will not be in force in the near future, coming up campaigns seem to be necessary to be promoted by the Deutscher

  13. Underreporting of maritime accidents to vessel accident databases.

    PubMed

    Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

    2011-11-01

    Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. PMID:21819835

  14. Columbia Accident Probe Widens

    NASA Technical Reports Server (NTRS)

    Covault, Craig

    2003-01-01

    The Columbia Accident Investigation Board has identified about a dozen shuttle program safety concerns it will address in its final report, in addition to foam shedding from the Lockheed Martin external tank-believed by many board members to be the direct cause for the loss of Columbia and her crew. As new evidence narrows the location of Columbia's left-wing breach to a lower corner of reinforced carbon-carbon (RCC) Panel 8 and its adjoining T-seal, the board is broadening its penetration of other shuttle safety issues. As the board works in Houston, United Space Alliance technicians here at Kennedy last week sent the first six of 22 RCC panels from the orbiter Atlantis left wing to Vought Aircraft Industries Inc. in Dallas for extensive testing to assess their integrity. The move is a key step toward both returning the shuttle to flight with Atlantis and obtaining more data on RCC panels subjected to fewer flights, and less exposure to the weather, than the older panels used on Columbia.

  15. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect

    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

  16. Aircraft accidents : method of analysis

    NASA Technical Reports Server (NTRS)

    1929-01-01

    This report on a method of analysis of aircraft accidents has been prepared by a special committee on the nomenclature, subdivision, and classification of aircraft accidents organized by the National Advisory Committee for Aeronautics in response to a request dated February 18, 1928, from the Air Coordination Committee consisting of the Assistant Secretaries for Aeronautics in the Departments of War, Navy, and Commerce. The work was undertaken in recognition of the difficulty of drawing correct conclusions from efforts to analyze and compare reports of aircraft accidents prepared by different organizations using different classifications and definitions. The air coordination committee's request was made "in order that practices used may henceforth conform to a standard and be universally comparable." the purpose of the special committee therefore was to prepare a basis for the classification and comparison of aircraft accidents, both civil and military. (author)

  17. Root causes and impacts of severe accidents at large nuclear power plants.

    PubMed

    Högberg, Lars

    2013-04-01

    The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities. PMID:23423737

  18. The Chernobyl accident ten years later

    SciTech Connect

    Squires, D.J.

    1996-12-31

    On April 26, 1986 at 1:23 AM a fire and explosion occurred at the fourth unit of the Chernobyl Nuclear Power Plant Complex, located in the Ukraine, that resulted in the destruction of the reactor core and most of the building in which it was housed. Several environmental impacts resulting from the accident will be discussed in this paper, which will include the effects on plant and wild life, radioactive waste generated and stored or disposed of, effects of evacuations relating to residents within the subsequently established 10km and 30km control zones, impacts of the emergency containment structure (sarcophagus), and potential effects on world opinion and future development of nuclear power. As an immediate result of the fire, 31 people died (2 from the fire & smoke, and 29 from excessive radiation); 237 cases of acute radiation sickness occurred; the total fatalities based upon induced chronic diseases as a result of the accident is unknown: more than 100,000 people were evacuated from within the subsequently established 30 km control zone; in excess of 50 million curies of radionuclides that included finely dispersed nuclear fuel, fragments of graphite, concrete and other building materials were released from the reactor into the environment; an estimated one million cubic meters of radioactive waste were generated (LLW, ILW, HLW); more than 5000 tons of materials (sand, boron, dolomite, cement, and lead) were used to put the fire out by helicopter; shutdown of the adjacent power plants were performed; and other environmental impacts occurred. The Chernobyl Nuclear Power Plant Unit No 4 is an RBMK-1000. It initiated operations in 1983, it was a 1000 MWe with a power output of 3200 MW(th), the reactor core contained 190 MT of fuel, with 1659 assemblies (plus 211 control rods), the average burnup rate was 10.3 MWd/kg, and the reactor operated on a continuous basis with maintenance and fuel reload performed during operations.

  19. 49 CFR 801.32 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Accident reports. 801.32 Section 801.32... PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.32 Accident reports. (a) The NTSB....S. civil transportation accidents, in accordance with 49 U.S.C. 1131(e). (b) These reports may...

  20. 29 CFR 1960.29 - Accident investigation.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 29 Labor 9 2011-07-01 2011-07-01 false Accident investigation. 1960.29 Section 1960.29 Labor... MATTERS Inspection and Abatement § 1960.29 Accident investigation. (a) While all accidents should be investigated, including accidents involving property damage only, the extent of such investigation shall...

  1. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 must, as soon...

  2. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 4 2011-10-01 2011-10-01 false Accident reports. 230.22 Section 230.22... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam... persons, the railroad on whose line the accident occurred shall immediately make a telephone report of...

  3. 49 CFR 845.40 - Accident report.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Accident report. 845.40 Section 845.40... RULES OF PRACTICE IN TRANSPORTATION; ACCIDENT/INCIDENT HEARINGS AND REPORTS Board Reports § 845.40 Accident report. (a) The Board will issue a detailed narrative accident report in connection with...

  4. 29 CFR 1960.29 - Accident investigation.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 9 2010-07-01 2010-07-01 false Accident investigation. 1960.29 Section 1960.29 Labor... MATTERS Inspection and Abatement § 1960.29 Accident investigation. (a) While all accidents should be investigated, including accidents involving property damage only, the extent of such investigation shall...

  5. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Accident reports. 230.22 Section 230.22... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam... persons, the railroad on whose line the accident occurred shall immediately make a telephone report of...

  6. 49 CFR 195.54 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 shall as soon...

  7. 49 CFR 845.40 - Accident report.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Accident report. 845.40 Section 845.40... RULES OF PRACTICE IN TRANSPORTATION; ACCIDENT/INCIDENT HEARINGS AND REPORTS Board Reports § 845.40 Accident report. (a) The Board will issue a detailed narrative accident report in connection with...

  8. 49 CFR 801.32 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Accident reports. 801.32 Section 801.32... PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.32 Accident reports. (a) The NTSB....S. civil transportation accidents, in accordance with 49 U.S.C. 1131(e). (b) These reports may...

  9. Spine Immobilizer for Accident Victims

    NASA Technical Reports Server (NTRS)

    Vykukal, H. C.; Lampson, K.

    1983-01-01

    Proposed conformal bladder filled with tiny spheres called "microballoons," enables spine of accident victim to be rapidly immobilized and restrained and permit victim to be safely removed from accident scene in extremely short time after help arrives. Microballoons expand to form rigid mass when pressure within bladder is less than ambient. Bladder strapped to victim is also strapped to rescue chair. Void between bladder and chair is filled with cloth wedges.

  10. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  11. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  12. Severe accident analysis using dynamic accident progression event trees

    NASA Astrophysics Data System (ADS)

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

  13. Heat up and potential failure of BWR upper internals during a severe accident

    SciTech Connect

    Robb, Kevin R

    2015-01-01

    In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

  14. Accident Tolerant Fuel Analysis

    SciTech Connect

    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

  15. Accident tolerant fuel analysis

    SciTech Connect

    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

  16. Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

    SciTech Connect

    Wichner, R.P.; Weber, C.F.; Hodge, S.A.; Beahm, E.C.; Wright, A.L.

    1984-01-01

    This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor.

  17. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  18. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    NASA Astrophysics Data System (ADS)

    Sabundjian, Gaianê; Andrade, Delvonei A.; Belchior, Antonio, Jr.; da Silva Rocha, Marcelo; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; de Souza Lima, Ana Cecília

    2013-05-01

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  19. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  20. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  1. [Venomous animal accidents in childhood

    PubMed

    Oliveira, J S; Campos, J A; Costa, D M

    1999-11-01

    OBJECTIVE: To highlight the importance of venomous animal accidents in childhood. The conducts are based on the proposals of the Ministério da Saúde do Brasil [Ministry of Health of Brazil] to standardize medical care in this kind of accident. This article shows the importance of early clinical diagnosis and assistance.METHODS: Review of international and national literature that includes original articles, official standards and books.RESULTS: Pediatricians may always feel insecure when they have to attend children who had venomous animal accidents because this kind of pathology is not very common. This article tries to offer easy guidelines and describes the main steps to be followed. Besides, peculiar or unusual aspects of these accidents are to be found in the literature referred to in the end of this article. Venomous animal accidents are always more severe in children, therefore resulting in higher mortality and sequelae. We assert that the early antivenom sera is extremely helpful.CONCLUSIONS: The systematization of the assistance may guarantee that the essential steps are followed thus making the assistance itself more effective. This is the purpose of the guidelines presented in this article. PMID:14685472

  2. Accident Prevention in the Cowshed*

    PubMed Central

    Mainzer, W.

    1966-01-01

    Work accidents were studied at two agricultural settlements in the Haifa area. Most of the accidents were caused by farm animals, particularly by cattle, a fact which is in agreement with a general statistical survey conducted by the Department for Occupational Health of the General Federation of Labour in Israel. However, in the present investigation it was found that the accident rate in cowsheds was more than 10 times higher among the members of a co-operative smallholders' village (Moshav Ovdim) than it was at a collective settlement (Moshav Shetufi) of the same numerical size. Searching for the basic factors involved, it was discovered that the main reason for this striking difference in accident frequency was the faulty design of the small cowshed at the individual farms of the co-operative settlement, which, lacking adequate protective measures, exposes the farmer to close contact with the animals throughout almost all stages of dealing with them. It is concluded that accidents from handling cattle can be prevented by adhering strictly to the principle of eliminating direct contact between man and animals in the construction of the cowshed and its annexes. Images PMID:5948271

  3. Nuclear accident dosimetry intercomparison studies.

    PubMed

    Sims, C S

    1989-09-01

    Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry. PMID:2777549

  4. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Cleveland, J.; Krueger, K.; Kernforschungsanlage Juelich G.m.b.H. . Arbeitsgemeinschaft Versuchsreaktor)

    1989-01-01

    Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have been conducted with the German pebble-bed High-Temperature Reactor AVR. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The tests demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into modular HTGR designs. The main LOCA test lasted for 5 d. After the test began, core temperatures increased for {approximately}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded accident levels of decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components. 3 refs., 9 figs., 1 tab.

  5. Characterization of plutonium particles originating from the BOMARC accident - 1960

    NASA Astrophysics Data System (ADS)

    Gostic, Richard Charles

    Within the U.S. arsenal, 32 accidents with nuclear weapons were reported between 1950 and 1980. One of these accidents occurred at McGuire AFB in 1960. A BOMARC missile armed with a nuclear warhead caught on fire and as a result the warhead was destroyed. Sub-millimeter particles consisting of weapons grade plutonium (WGPu) produced by this accident were distributed around the site and remained in the environment for 47 years. Soil cores known to contain WGPu particles produced by this accident were obtained. The particles were localized and removed from the soil with the aid of high resolution computed tomography. The isotopic composition of the particles and the date of manufacture of the Pu were estimated using a combination of alpha and gamma spectroscopy. Scanning electron microscopy was used to study the surface morphology of the particles; energy dispersive spectroscopy and synchrotron based x-ray fluorescence were used to determine the composition and elemental distributions of the particles. The results of these experiments and their application to the field of nuclear forensic analysis are discussed in this thesis.

  6. Apparatus for controlling nuclear core debris

    DOEpatents

    Jones, Robert D.

    1978-01-01

    Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

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

    SciTech Connect

    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.

  8. Severe-accident-sequence assessment of hypothetical complete-station blackout at the Browns Ferry Nuclear Plant

    SciTech Connect

    Yue, D.D.; Condon, W.A.

    1981-01-01

    An investigation has been made of various accident sequence which may occur following a complete loss of offsite and onsite ac power at a Boiling Water Reactor nuclear power plant. The investigation was performed for the Browns Ferry Nuclear Power Plant, and all accident sequences resulted in a hypothetical core meltdown. Detailed calculations were performed with the MARCH computer meltdown. Detailed calcuations were performed with the MARCH computer code containing a decay power calculation which was modified to include the actinides. This change has resulted in shortening the time before core uncovery by approx. 18%, and reducing the time before the start of core melting by approx. 26%. Following the hypothetical core meltdown accident, the drywell electric penetration assembly seals have been identified as the most likely leak pathway outside the containment. This potential mode of containment failure occurs at a pressure approx. 30% lower than that analyzed in the Reactor Safety Study.

  9. [Diving accidents. Emergency treatment of serious diving accidents].

    PubMed

    Schröder, S; Lier, H; Wiese, S

    2004-11-01

    Decompression injuries are potentially life-threatening incidents mainly due to a rapid decline in ambient pressure. Decompression illness (DCI) results from the presence of gas bubbles in the blood and tissue. DCI may be classified as decompression sickness (DCS) generated from the liberation of gas bubbles following an oversaturation of tissues with inert gas and arterial gas embolism (AGE) mainly due to pulmonary barotrauma. People working under hyperbaric pressure, e.g. in a caisson for general construction under water, and scuba divers are exposed to certain risks. Diving accidents can be fatal and are often characterized by organ dysfunction, especially neurological deficits. They have become comparatively rare among professional divers and workers. However, since recreational scuba diving is gaining more and more popularity there is an increasing likelihood of severe diving accidents. Thus, emergency staff working close to areas with a high scuba diving activity, e.g. lakes or rivers, may be called more frequently to a scuba diving accident. The correct and professional emergency treatment on site, especially the immediate and continuous administration of normobaric oxygen, is decisive for the outcome of the accident victim. The definitive treatment includes rapid recompression with hyperbaric oxygen. The value of adjunctive medication, however, remains controversial. PMID:15565421

  10. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. )

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.

  11. 49 CFR 229.17 - Accident reports.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... CFR part 225. ... 49 Transportation 4 2011-10-01 2011-10-01 false Accident reports. 229.17 Section 229.17..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS General § 229.17 Accident reports. (a)...

  12. 49 CFR 229.17 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... CFR part 225. ... 49 Transportation 4 2010-10-01 2010-10-01 false Accident reports. 229.17 Section 229.17..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS General § 229.17 Accident reports. (a)...

  13. [What kind of accidents can happen in a nuclear power plant].

    PubMed

    Debes, M

    1995-03-15

    The lessons drawn from real reactor accidents are of great value. The safety approach in France relies on defence in depth and takes into account accidents in the plant design, completed by a probabilistic approach and experience feedback. Ultimate procedures are implemented on the basis of severe accidents studies which include core melting or partial containment defect, in order to mitigate their consequences even if they are improbable, and to enable a proper implementation of emergency planning countermeasures. The accident hypothesis and consequences are considered to draw the emergency planning procedures. Off site countermeasures, such as in house-confinement, limited evacuation or iodine distribution, are efficient in limiting the consequences for the public. Experience feedback, in association with a proactive vigilance and prevention policy, is developed in order to detect and correct in a proactive way the root causes of any deviation, even minor, so as to avoid multiple failures and ensure safety. PMID:7754324

  14. Markov Model of Severe Accident Progression and Management

    SciTech Connect

    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.

  15. The Physics of Traffic Accidents

    ERIC Educational Resources Information Center

    Knight, Peter

    1975-01-01

    Shows how physics can be used to analyze and prevent traffic accidents by determining critical speeds on curves, the behavior of motor cycles and stability of articulated vehicles, and the visibility that is needed to make a minor road junction safe. (MLH)

  16. Delta launch vehicle accident investigation

    NASA Astrophysics Data System (ADS)

    1986-03-01

    The text of the testimony given by several witnesses during the House hearings on the Delta launch vehicle accident of May 3, 1986 is given. Pre-launch procedures, failure analysis, the possibility of sabotage, and design and testing are among the topics discussed.

  17. Updating outdated predictive accident models.

    PubMed

    Wood, A G; Mountain, L J; Connors, R D; Maher, M J; Ropkins, K

    2013-06-01

    Reliable predictive accident models (PAMs) (also referred to as safety performance functions (SPFs)) are essential to design and maintain safe road networks however, ongoing changes in road and vehicle design coupled with road safety initiatives, mean that these models can quickly become dated. Unfortunately, because the fitting of sophisticated PAMs including a wide range of explanatory variables is not a trivial task, available models tend to be based on data collected many years ago and seem unlikely to give reliable estimates of current accidents. Large, expensive studies to produce new models are likely to be, at best, only a temporary solution. This paper thus seeks to develop a practical and efficient methodology to allow currently available PAMs to be updated to give unbiased estimates of accident frequencies at any point in time. Two principal issues are examined: the extent to which the temporal transferability of predictive accident models varies with model complexity; and the practicality and efficiency of two alternative updating strategies. The models used to illustrate these issues are the suites of models developed for rural dual and single carriageway roads in the UK. These are widely used in several software packages in spite of being based on data collected during the 1980s and early 1990s. It was found that increased model complexity by no means ensures better temporal transferability and that calibration of the models using a scale factor can be a practical alternative to fitting new models. PMID:23510788

  18. Time Slows Down during Accidents

    PubMed Central

    Arstila, Valtteri

    2012-01-01

    The experienced speed of the passage of time is not constant as time can seem to fly or slow down depending on the circumstances we are in. Anecdotally accidents and other frightening events are extreme examples of the latter; people who have survived accidents often report altered phenomenology including how everything appeared to happen in slow motion. While the experienced phenomenology has been investigated, there are no explanations about how one can have these experiences. Instead, the only recently discussed explanation suggests that the anecdotal phenomenology is due to memory effects and hence not really experienced during the accidents. The purpose of this article is (i) to reintroduce the currently forgotten comprehensively altered phenomenology that some people experience during the accidents, (ii) to explain why the recent experiments fail to address the issue at hand, and (iii) to suggest a new framework to explain what happens when people report having experiences of time slowing down in these cases. According to the suggested framework, our cognitive processes become rapidly enhanced. As a result, the relation between the temporal properties of events in the external world and in internal states becomes distorted with the consequence of external world appearing to slow down. That is, the presented solution is a realist one in a sense that it maintains that sometimes people really do have experiences of time slowing down. PMID:22754544

  19. Time Slows Down during Accidents.

    PubMed

    Arstila, Valtteri

    2012-01-01

    The experienced speed of the passage of time is not constant as time can seem to fly or slow down depending on the circumstances we are in. Anecdotally accidents and other frightening events are extreme examples of the latter; people who have survived accidents often report altered phenomenology including how everything appeared to happen in slow motion. While the experienced phenomenology has been investigated, there are no explanations about how one can have these experiences. Instead, the only recently discussed explanation suggests that the anecdotal phenomenology is due to memory effects and hence not really experienced during the accidents. The purpose of this article is (i) to reintroduce the currently forgotten comprehensively altered phenomenology that some people experience during the accidents, (ii) to explain why the recent experiments fail to address the issue at hand, and (iii) to suggest a new framework to explain what happens when people report having experiences of time slowing down in these cases. According to the suggested framework, our cognitive processes become rapidly enhanced. As a result, the relation between the temporal properties of events in the external world and in internal states becomes distorted with the consequence of external world appearing to slow down. That is, the presented solution is a realist one in a sense that it maintains that sometimes people really do have experiences of time slowing down. PMID:22754544

  20. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    SciTech Connect

    Siefken, L.J.

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ``Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

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

    SciTech Connect

    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.

  2. Determinants of injuries in passenger vessel accidents.

    PubMed

    Yip, Tsz Leung; Jin, Di; Talley, Wayne K

    2015-09-01

    This paper investigates determinants of crew and passenger injuries in passenger vessel accidents. Crew and passenger injury equations are estimated for ferry, ocean cruise, and river cruise vessel accidents, utilizing detailed data of individual vessel accidents that were investigated by the U.S. Coast Guard during the time period 2001-2008. The estimation results provide empirical evidence (for the first time in the literature) that crew injuries are determinants of passenger injuries in passenger vessel accidents. PMID:26070017

  3. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    SciTech Connect

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-12-23

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  4. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    NASA Astrophysics Data System (ADS)

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-12-01

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  5. 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

    Whitehead, D.; Darby, J.; Yakle, J.

    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.

  6. School Bus Accidents and Driver Age.

    ERIC Educational Resources Information Center

    McMichael, Judith

    The study examines the rates and types of school bus accidents according to the age of the school bus driver. Accident rates in North Carolina for the school year 1971-72 were analyzed using three sources of data: accident reports, driver and mileage data, and questionnaires administered to a sample of school bus drivers. Data were obtained on…

  7. 48 CFR 836.513 - Accident prevention.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Accident prevention. 836... CATEGORIES OF CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 836.513 Accident prevention. The contracting officer must insert the clause at 852.236-87, Accident Prevention,...

  8. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 3 2011-10-01 2011-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  9. 48 CFR 1836.513 - Accident prevention.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 6 2011-10-01 2011-10-01 false Accident prevention. 1836... 1836.513 Accident prevention. The contracting officer must insert the clause at 1852.223-70, Safety and Health, in lieu of FAR clause 52.236-13, Accident Prevention, and its Alternate I....

  10. 48 CFR 36.513 - Accident prevention.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 1 2011-10-01 2011-10-01 false Accident prevention. 36... CATEGORIES OF CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 36.513 Accident prevention. (a) The contracting officer shall insert the clause at 52.236-13, Accident Prevention,...

  11. 48 CFR 636.513 - Accident prevention.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 4 2011-10-01 2011-10-01 false Accident prevention. 636... CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 636.513 Accident prevention. (a) In... contracting activities shall insert DOSAR 652.236-70, Accident Prevention, in lieu of FAR clause...

  12. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... Foreign Relations DEPARTMENT OF STATE ECONOMIC AND OTHER FUNCTIONS CIVIL AVIATION United States Aircraft Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration... accident to the Foreign Service post, to the nearest Civil Aeronautics Administration office, and to...

  13. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... Foreign Relations DEPARTMENT OF STATE ECONOMIC AND OTHER FUNCTIONS CIVIL AVIATION United States Aircraft Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration... accident to the Foreign Service post, to the nearest Civil Aeronautics Administration office, and to...

  14. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 22 Foreign Relations 1 2011-04-01 2011-04-01 false Reporting accidents. 102.8 Section 102.8... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration... and the supervisory Foreign Service office whenever considered appropriate. A final report, after...

  15. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 28 Judicial Administration 2 2011-07-01 2011-07-01 false Repetitious accidents. 301.106 Section 301.106 Judicial Administration FEDERAL PRISON INDUSTRIES, INC., DEPARTMENT OF JUSTICE INMATE ACCIDENT COMPENSATION General § 301.106 Repetitious accidents. If an inmate worker is involved in successive...

  16. 48 CFR 836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Accident prevention. 836... prevention. The contracting officer must insert the clause at 852.236-87, Accident Prevention, in solicitations and contracts for construction that contain the clause at FAR 52.236-13, Accident Prevention....

  17. 48 CFR 1836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Accident prevention. 1836... 1836.513 Accident prevention. The contracting officer must insert the clause at 1852.223-70, Safety and Health, in lieu of FAR clause 52.236-13, Accident Prevention, and its Alternate I....

  18. NASA Medical Response to Human Spacecraft Accidents

    NASA Technical Reports Server (NTRS)

    Patlach, Robert

    2010-01-01

    Manned space flight is risky business. Accidents have occurred and may occur in the future. NASA's manned space flight programs, with all their successes, have had three fatal accidents, one at the launch pad and two in flight. The Apollo fire and the Challenger and Columbia accidents resulted in a loss of seventeen crewmembers. Russia's manned space flight programs have had three fatal accidents, one ground-based and two in flight. These accidents resulted in the loss of five crewmembers. Additionally, manned spacecraft have encountered numerous close calls with potential for disaster. The NASA Johnson Space Center Flight Safety Office has documented more than 70 spacecraft incidents, many of which could have become serious accidents. At the Johnson Space Center (JSC), medical contingency personnel are assigned to a Mishap Investigation Team. The team deploys to the accident site to gather and preserve evidence for the Accident Investigation Board. The JSC Medical Operations Branch has developed a flight surgeon accident response training class to capture the lessons learned from the Columbia accident. This presentation will address the NASA Mishap Investigation Team's medical objectives, planned response, and potential issues that could arise subsequent to a manned spacecraft accident. Educational Objectives are to understand the medical objectives and issues confronting the Mishap Investigation Team medical personnel subsequent to a human space flight accident.

  19. 49 CFR 195.50 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An...

  20. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Repetitious accidents. 301.106 Section 301.106 Judicial Administration FEDERAL PRISON INDUSTRIES, INC., DEPARTMENT OF JUSTICE INMATE ACCIDENT COMPENSATION General § 301.106 Repetitious accidents. If an inmate worker is involved in successive...

  1. Radiation protection: an analysis of thyroid blocking. [Effectiveness of KI in reducing radioactive uptake following potential reactor accident

    SciTech Connect

    Aldrich, D C; Blond, R M

    1980-01-01

    An analysis was performed to provide guidance to policymakers concerning the effectiveness of potassium iodide (KI) as a thyroid blocking agent in potential reactor accident situations, the distance to which (or area within which) it should be distributed, and its relative effectiveness compared to other available protective measures. The analysis was performed using the Reactor Safety Study (WASH-1400) consequence model. Four categories of accidents were addressed: gap activity release accident (GAP), GAP without containment isolation, core melt with a melt-through release, and core melt with an atmospheric release. Cost-benefit ratios (US $/thyroid nodule prevented) are given assuming that no other protective measures are taken. Uncertainties due to health effects parameters, accident probabilities, and costs are assessed. The effects of other potential protective measures, such as evacuation and sheltering, and the impact on children (critical population) are evaluated. Finally, risk-benefit considerations are briefly discussed.

  2. [Case report: fatal diving-accident. Or: accident while diving?].

    PubMed

    Böttcher, F; Jüttner, B; Krause, A; Rocha, M; Koppert, W

    2012-02-01

    This example of a fatal diving accident shows how challenging such cases can be in pre-hospital and clinical care. There is no common mechanism in diving fatalities and more than one group of disorders coming along with decompression sickness. Diving medicine is not an element of medical education, which results in insecurity and hampers adequate therapy of diving incidents. This is aggravated by an insufficient availability of hyperbaric chambers in Germany. PMID:22354401

  3. The Chornobyl Accident: A Comprehensive Risk Assessment

    SciTech Connect

    Poyarkov, Victor A.; Vargo, George J.; George J. Vargo

    2000-01-01

    This book provides a comprehensive of the April 1986 Chornobyl Nuclear Power Plant accident and its short and long-term effects in the fourteen years since the accident. Chapters include: cause and description of the accident; the Shelter constructed to contain the remains the destroyed reactor, radioactive wastes arising from the accident, environmental contamination, individual and collective radiation doses, societal aspects, economic impact and conclusions. Appendices on radiological units, the medical consequences of the accident, and a list of acronym and abbreviations are included.

  4. Single pilot IFR accident data analysis

    NASA Technical Reports Server (NTRS)

    Harris, D. F.

    1983-01-01

    The aircraft accident data recorded by the National Transportation and Safety Board (NTSR) for 1964-1979 were analyzed to determine what problems exist in the general aviation (GA) single pilot instrument flight rule (SPIFR) environment. A previous study conducted in 1978 for the years 1964-1975 provided a basis for comparison. This effort was generally limited to SPIFR pilot error landing phase accidents but includes some SPIFR takeoff and enroute accident analysis as well as some dual pilot IFR accident analysis for comparison. Analysis was performed for 554 accidents of which 39% (216) occurred during the years 1976-1979.

  5. Severe accident sequence assessment for boiling water reactors: program overview

    SciTech Connect

    Fontana, M. H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case.

  6. Accidents in the aluminium smelting industry.

    PubMed

    Das, B C; Chaudhury, S

    1995-01-01

    Analysis of the accident records of an aluminium smelting industry, covering about 2,100 employees, over a period of three years, showed a total of 465 accidents of male employees. Out of these, 5 were fatal, 40.86% were from contacts with extreme temperatures, causing burn injury to 42.58%. Hot materials were the agents causing 44.52% of the burn injuries. Molten aluminium constituted 43.96% amongst hot materials. Injury to lower limbs constituted 38.71% and that to upper limbs 36.99%. The accidents occurring to the employees, in the age group of 26-33 years, amounted to 61.72% of the total accidents. The average number of man-days lost per year was 11,153. Average frequency rate of accidents was 30.75 accidents per million man-hours worked. Severity rate of accidents was 2.196 per million man-hours worked. Incident rate per thousand employees was 73.81. Average number of days lost per accidents was 71.95 days and average duration of man-hours between accidents was 32,516. Mean age of the employees, who met with the accidents were 29.53 years. Share of accidents in the second half of each shift was always more than that in the first half, and this average was 66.66%. PMID:8557540

  7. The Concept of Accident Proneness: A Review

    PubMed Central

    Froggatt, Peter; Smiley, James A.

    1964-01-01

    The term accident proneness was coined by psychological research workers in 1926. Since then its concept—that certain individuals are always more likely than others to sustain accidents, even though exposed to equal risk—has been questioned but seldom seriously challenged. This article describes much of the work and theory on which this concept is based, details the difficulties encountered in obtaining valid information and the interpretative errors that can arise from the examination of imperfect data, and explains why accident proneness became so readily accepted as an explanation of the facts. A recent hypothesis of accident causation, namely that a person's accident liability may vary from time to time, is outlined, and the respective abilities of this and of accident proneness to accord with data from the more reliable literature are examined. The authors conclude that the hypothesis of individual variation in liability is more realistic and in better agreement with the data than is accident proneness. PMID:14106130

  8. Accident/Mishap Investigation System

    NASA Technical Reports Server (NTRS)

    Keller, Richard; Wolfe, Shawn; Gawdiak, Yuri; Carvalho, Robert; Panontin, Tina; Williams, James; Sturken, Ian

    2007-01-01

    InvestigationOrganizer (IO) is a Web-based collaborative information system that integrates the generic functionality of a database, a document repository, a semantic hypermedia browser, and a rule-based inference system with specialized modeling and visualization functionality to support accident/mishap investigation teams. This accessible, online structure is designed to support investigators by allowing them to make explicit, shared, and meaningful links among evidence, causal models, findings, and recommendations.

  9. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  10. Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

    SciTech Connect

    Roald A. Wigeland; James E. Cahalan

    2011-04-01

    Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

  11. Preparation, conduct, and experimental results of the AVR loss-of-coolant accident simulation test

    SciTech Connect

    Kruger, K.; Bergerfurth, A.; Burger, S.; Pohl, P.; Wimmers, M. ); Cleveland, J.C. )

    1991-02-01

    A loss-of-coolant accident (LOCA) is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small high-temperature gas-cooled reactor (HTGR) design, LOCA simulation tests have been conducted at the Arbeitsgemeinschaft Versuchsreaktor (AVR), the German pebble-bed-high-temperature reactor plant. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions without emergency cooling. This paper presents the planning and licensing activities including pretest predictions performed for the LOCA test are described, and the conduct of the test and experimental results. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The test demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into optimized modular HTGR designs. The main LOCA test lasted for 5 days. After the test began, core temperatures increased for {approx}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded the simulated decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components.

  12. Dual-core antiresonant hollow core fibers.

    PubMed

    Liu, Xuesong; Fan, Zhongwei; Shi, Zhaohui; Ma, Yunfeng; Yu, Jin; Zhang, Jing

    2016-07-25

    In this work, dual-core antiresonant hollow core fibers (AR-HCFs) are numerically demonstrated, based on our knowledge, for the first time. Two fiber structures are proposed. One is a composite of two single-core nested nodeless AR-HCFs, exhibiting low confinement loss and a circular mode profile in each core. The other has a relatively simple structure, with a whole elliptical outer jacket, presenting a uniform and wide transmission band. The modal couplings of the dual-core AR-HCFs rely on a unique mechanism that transfers power through the air. The core separation and the gap between the two cores influence the modal coupling strength. With proper designs, both of the dual-core fibers can have low phase birefringence and short modal coupling lengths of several centimeters. PMID:27464191

  13. Defining a common set of indicators to monitor road accidents in the European Union

    PubMed Central

    Farchi, Sara; Molino, Nunzio; Giorgi Rossi, Paolo; Borgia, Piero; Krzyzanowski, Michael; Dalbokova, Dafina; Kim, Rokho

    2006-01-01

    Background currently road accidents are mostly monitored through mortality and injury rates. This paper reports the methodology and the results of a project set forth by the European Union (EU) and coordinated by the WHO aimed at identifying and evaluating a core set of indicators to monitor the causal chain of road accident health effects. The project is part of the ECOEHIS (Development of Environment and Health Indicators for European Union Countries). Methods a group of experts (WG), identified 14 indicators after a review of the information collected at the EU level, each of them representing a specific aspect of the DPSEEA (Driving, Pressure, State, Exposure, Effect, Action) model applied and adapted to the road accidents. Each indicator was scored according to a list of 16 criteria chosen by the WG. Those found to have a high score were analysed to determine if they were compatible with EU legislation and then tested in the feasibility study. Results 11 of the 14 indicators found to be relevant and compatible with the criteria of selection were proposed for the feasibility study. Mortality, injury, road accident rate, age of vehicle fleet, and distance travelled are the indicators recommended for immediate implementation. Conclusion after overcoming the limitations that emerged (absence of a common definition of death by road accident and injury severity, underestimation of injuries, differences in information quality) this core set of indicators will allow Member States to carry out effective internal/external comparisons over time. PMID:16834780

  14. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab.

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

    SciTech Connect

    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.

  16. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect

    Wang, S.-J.; Chiang, K.-S.; Chiang, S.-C.

    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.

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

    SciTech Connect

    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)

  18. The role of chemical reactions in the Chernobyl accident

    SciTech Connect

    Grishanin, E. I.

    2010-12-15

    It is shown that chemical reactions played an essential role in the Chernobyl accident at all of its stages. It is important that the reactor before the explosion was at maximal xenon poisoning, and its reactivity, apparently, was not destroyed by the explosion. The reactivity release due to decay of Xe-235 on the second day after the explosion led to a reactor power of 80-110 MW. Owing to this power, the chemical reactions of reduction of uranium, plutonium, and other metals at a temperature of about 2000 Degree-Sign C occurred in the core. The yield of fission products thus sharply increased. Uranium and other metals flew down in the bottom water communications and rooms. After reduction of the uranium and its separation from the graphite, the chain reaction stopped, the temperature of the core decreased, and the activity yield stopped.

  19. Code System for the Analysis of Material Test Reactor (MTR) Cores.

    Energy Science and Technology Software Center (ESTSC)

    1995-03-24

    Version 00 The RETRAC code uses a set of coupled neutron point-kinetics equations and thermal-hydraulic conservation laws to simulate nuclear reactor core behavior under transient or accident conditions. The reactor core is represented by a single equivalent unit cell composed of three regions: fuel, clad, and moderator (coolant).

  20. Temporal Statistic of Traffic Accidents in Turkey

    NASA Astrophysics Data System (ADS)

    Erdogan, S.; Yalcin, M.; Yilmaz, M.; Korkmaz Takim, A.

    2015-10-01

    Traffic accidents form clusters in terms of geographic space and over time which themselves exhibit distinct spatial and temporal patterns. There is an imperative need to understand how, where and when traffic accidents occur in order to develop appropriate accident reduction strategies. An improved understanding of the location, time and reasons for traffic accidents makes a significant contribution to preventing them. Traffic accident occurrences have been extensively studied from different spatial and temporal points of view using a variety of methodological approaches. In literature, less research has been dedicated to the temporal patterns of traffic accidents. In this paper, the numbers of traffic accidents are normalized according to the traffic volume and the distribution and fluctuation of these accidents is examined in terms of Islamic time intervals. The daily activities and worship of Muslims are arranged according to these time intervals that are spaced fairly throughout the day according to the position of the sun. The Islamic time intervals are never been used before to identify the critical hour for traffic accidents in the world. The results show that the sunrise is the critical time that acts as a threshold in the rate of traffic accidents throughout Turkey in Islamic time intervals.

  1. Risk and protection factors in fatal accidents.

    PubMed

    Dupont, Emmanuelle; Martensen, Heike; Papadimitriou, Eleonora; Yannis, George

    2010-03-01

    This paper aims at addressing the interest and appropriateness of performing accident severity analyses that are limited to fatal accident data. Two methodological issues are specifically discussed, namely the accident-size factors (the number of vehicles in the accident and their level of occupancy) and the comparability of the baseline risk. It is argued that - although these two issues are generally at play in accident severity analyses - their effects on, e.g., the estimation of survival probability, are exacerbated if the analysis is limited to fatal accident data. As a solution, it is recommended to control for these effects by (1) including accident-size indicators in the model, (2) focusing on different sub-groups of road-users while specifying the type of opponent in the model, so as to ensure that comparable baseline risks are worked with. These recommendations are applied in order to investigate risk and protection factors of car occupants involved in fatal accidents using data from a recently set up European Fatal Accident Investigation database (Reed and Morris, 2009). The results confirm that the estimated survival probability is affected by accident-size factors and by type of opponent. The car occupants' survival chances are negatively associated with their own age and that of their vehicle. The survival chances are also lower when seatbelt is not used. Front damage, as compared to other damaged car areas, appears to be associated with increased survival probability, but mostly in the case in which the accident opponent was another car. The interest of further investigating accident-size factors and opponent effects in fatal accidents is discussed. PMID:20159090

  2. "What--me worry?" "Why so serious?": a personal view on the Fukushima nuclear reactor accidents.

    PubMed

    Gallucci, Raymond

    2012-09-01

    Infrequently, it seems that a significant accident precursor or, worse, an actual accident, involving a commercial nuclear power reactor occurs to remind us of the need to reexamine the safety of this important electrical power technology from a risk perspective. Twenty-five years since the major core damage accident at Chernobyl in the Ukraine, the Fukushima reactor complex in Japan experienced multiple core damages as a result of an earthquake-induced tsunami beyond either the earthquake or tsunami design basis for the site. Although the tsunami itself killed tens of thousands of people and left the area devastated and virtually uninhabitable, much concern still arose from the potential radioactive releases from the damaged reactors, even though there was little population left in the area to be affected. As a lifelong probabilistic safety analyst in nuclear engineering, even I must admit to a recurrence of the doubt regarding nuclear power safety after Fukushima that I had experienced after Three Mile Island and Chernobyl. This article is my attempt to "recover" my personal perspective on acceptable risk by examining both the domestic and worldwide history of commercial nuclear power plant accidents and attempting to quantify the risk in terms of the frequency of core damage that one might glean from a review of operational history. PMID:22394214

  3. Continuously improving safety of nuclear installations: An approach to be reinforced after the Fukushima accident

    NASA Astrophysics Data System (ADS)

    Repussard, Jacques; Schwarz, Michel

    2012-05-01

    After the Three Mile Island accident in 1979 and the Chernobyl accident in 1986, the Fukushima accident shows that the probability of a core meltdown accident in an LWR (Light Water Reactor) has been largely underestimated. The consequences of such an accident are unacceptable: except in the case of TMI2 (Three Mile Island 2) large areas around the damaged plants are contaminated for decades and populations have to be relocated for long periods. This article presents the French approach which consists in improving continuously the safety of the Nuclear Power Plants (NPP) on the basis of lessons learned from operating experience and from the progress in R&D (Research and Development). It details the key role played by IRSN (Institut de radioprotection et de sûreté nucléaire), the French TSO (Technical and scientific Safety Organization), and shows how the Fukushima accident contributes to this approach in improving NPP robustness. It concludes on the necessity to keep on networking TSOs, to share knowledge as well as R&D resources, with the ultimate goal of enhancing and harmonizing nuclear safety worldwide.

  4. MELCOR code analysis of a severe accident LOCA at Peach Bottom Plant

    SciTech Connect

    Carbajo, J.J. )

    1993-01-01

    A design-basis loss-of-coolant accident (LOCA) concurrent with complete loss of the emergency core cooling systems (ECCSs) has been analyzed for the Peach Bottom atomic station unit 2 using the MELCOR code, version 1.8.1. The purpose of this analysis is to calculate best-estimate times for the important events of this accident sequence and best-estimate source terms. Calculated pressures and temperatures at the beginning of the transient have been compared to results from the Peach Bottom final safety analysis report (FSAR). MELCOR-calculated source terms have been compared to source terms reported in the NUREG-1465 draft.

  5. Passive decay heat removal by natural air convection after severe accidents

    SciTech Connect

    Erbacher, F.J.; Neitzel, H.J.; Cheng, X.

    1995-09-01

    The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

  6. Radioactivity inspection of Taiwan for food products imported from Japan after the Fukushima nuclear accident.

    PubMed

    Chiu, Huang-Sheng; Huang, Ping-Ji; Wuu, Jyi-Lan; Wang, Jeng-Jong

    2013-11-01

    The 3-11 Earthquake occurred in Japan last year had greatly damaged the lives and properties and also caused the core meltdown accident in the Fukushima nuclear power plant followed by the leakage of radioactive materials into biosphere. In order to protect against the detriment of radiation from foods which were imported from Japan, the Institute of Nuclear Energy Research (INER) in Taiwan started to conduct radioactivity inspection of food products from Japan after the accident. A total of about 20,000 samples had been tested from March 24 2011 to March 31 2012. PMID:23583088

  7. MSHA releases data on CM crushing accidents

    SciTech Connect

    2007-02-15

    The US Mine Safety and Health Administration (MHSA) recently formed a committee to identify norms and trends in remote control continuous miner crushing accidents. The final report found that these types of accidents commonly happen to experienced miners during routine mining activities, with the majority occurring while moving the miner from one face to another, place changing. Another common aspect of the accidents is that many of the victims are experienced miners who are newly employed at the mine where the accident occurred. Training all employees to stay outside the turning radius of an energized remote control continuous miner, establishing this as a safe operating procedure, and consistently enforcing this practice among miners will reduce these types of accidents. This article was excerpted from the 'Remote Control Continuous Mining Machine Crushing Accident Data Study' published in May 2006. The report may be found from the website: www.msha.gov. 4 figs., 1 tab.

  8. Organisational accidents investigation methodology and lessons learned.

    PubMed

    Dien, Yves; Llory, Michel; Montmayeul, René

    2004-07-26

    The purpose of this paper is to reflect on accident analysis methods. As the understanding of industrial accidents and incidents has evolved, they are no longer considered as the sole product of human and/or technical failures but also as originating in an unfavourable organisational context. After presenting some theoretical developments which are responsible for this evolution, we will propose two examples of organisational accidents and incidents. We will then present some properties of organisational accidents, and we will focus on some "accident-generating" organisational factors. The definition of these factors comes from an empirical approach to event analysis. Finally, we will briefly present their implications for accident and incident analysis. PMID:15231360

  9. Transportation accident scenarios for commercial spent fuel

    SciTech Connect

    Wilmot, E L

    1981-02-01

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

  10. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    SciTech Connect

    S.O. Bader

    1999-10-18

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

  11. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect

    Ryan, G.W., Westinghouse Hanford

    1996-07-12

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  12. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  13. Industrial Safety and Accidents Prevention

    SciTech Connect

    Sajjad Akbar

    2006-07-01

    Accident Hazards, dangers, losses and risk are what we would to like to eliminate, minimize or avoid in industry. Modern industries have created many opportunities for these against which man's primitive instincts offer no protection. In today's complex industrial environment safety has become major preoccupation, especially after the realization that there is a clear economic incentive to do so. Industrial hazards may cause by human error or by physical or mechanical malfunction, it is very often possible to eliminate the worst consequences of human error by engineering modification. But the modification also needs checking very thoroughly to ensue that it has not introduced some new and unsuspected hazard. (author)

  14. Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1.

    PubMed

    Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

    2015-12-01

    In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by (137)Cesium ((137)Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as (132)Te-(132)I, (131)I, (134)Cs and (137)Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h(-1) per initial (137)Cs deposition of 1000 kBq m(-2), whereas it was 100 μGy h(-1) around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m(-2) for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums ((134)Cs + (137)Cs) around Chernobyl and Fukushima-1, respectively. PMID:26568603

  15. Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study

    PubMed Central

    AMIRI, Mehran; ARDESHIR, Abdollah; FAZEL ZARANDI, Mohammad Hossein

    2014-01-01

    Abstract Background The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. Methods The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were defined. The core of this work is devoted to analyzing the data from different perspectives such as age of workers, occupation and construction phase, day of the week, time of the day, seasonal analysis, regional considerations, type of accident, and body parts affected. Results Workers between 15-19 years old (TAR=13.4%) are almost six times more exposed to risk of accident than the average of all ages (TAR=2.51%). Laborers and structural workers (TAR=66.6%) and those working at heights (TAR=47.2%) experience more accidents than other groups of workers. Moreover, older workers over 65 years old (TSI=1.97%> average TSI=1.60%), work supervisors (TSI=12.20% >average TSI=9.09%), and night shift workers (TSI=1.89% >average TSI=1.47%) are more prone to severe accidents. Conclusion It is recommended that laborers, young workers, weekend and night shift workers be supervised more carefully in the workplace. Use of Personal Protective Equipment (PPE) should be compulsory in working environments, and special attention should be undertaken to people working outdoors and at heights. It is also suggested that policymakers pay more attention to the improvement of safety conditions in deprived and cold western regions. PMID:26005662

  16. Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1

    PubMed Central

    Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

    2015-01-01

    In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by 137Cesium (137Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as 132Te-132I, 131I, 134Cs and 137Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h−1 per initial 137Cs deposition of 1000 kBq m−2, whereas it was 100 μGy h−1 around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m−2 for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums (134Cs + 137Cs) around Chernobyl and Fukushima-1, respectively. PMID:26568603

  17. Aircraft Loss-of-Control Accident Analysis

    NASA Technical Reports Server (NTRS)

    Belcastro, Christine M.; Foster, John V.

    2010-01-01

    Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. To gain a better understanding into aircraft loss-of-control events and possible intervention strategies, this paper presents a detailed analysis of loss-of-control accident data (predominantly from Part 121), including worst case combinations of causal and contributing factors and their sequencing. Future potential risks are also considered.

  18. The determinants of fishing vessel accident severity.

    PubMed

    Jin, Di

    2014-05-01

    The study examines the determinants of fishing vessel accident severity in the Northeastern United States using vessel accident data from the U.S. Coast Guard for 2001-2008. Vessel damage and crew injury severity equations were estimated separately utilizing the ordered probit model. The results suggest that fishing vessel accident severity is significantly affected by several types of accidents. Vessel damage severity is positively associated with loss of stability, sinking, daytime wind speed, vessel age, and distance to shore. Vessel damage severity is negatively associated with vessel size and daytime sea level pressure. Crew injury severity is also positively related to the loss of vessel stability and sinking. PMID:24473412

  19. Accidents in family forestry's firewood production.

    PubMed

    Lindroos, Ola; Aspman, Emma Wilhelmson; Lidestav, Gun; Neely, Gregory

    2008-05-01

    Firewood is commonly used around the world, but little is known about the work involved in its production and associated accidents. The objectives were to identify relationships between accidents and time exposure, workers' age and sex, equipment used and work activities in family forestry's firewood production. Data from a postal survey in Northern Sweden were compared to a database of injuries in the same region. Most accidents occurred to 50-69 year old men, who also worked most hours. No significant differences in sex and age were found between expected and recorded accident frequencies when calculated from total work hours; however, when calculated using numbers of active persons significant differences were found for both age and sex. Frequency of accidents per unit worked time was higher for machine involving activities than for other activities. Accidents that occurred when using wedge splitter machines were responsible for most of this overrepresentation. Fingers were the most commonly injured body parts. Mean accident rate for the equipment used was 87 accidents per million work hours, and the rate was highest for wedge splitters (122 accidents per million work hours). Exposure to elevated risks due to violation of safety procedures is discussed, as well as possible preventative measures. PMID:18460354

  20. Occupational Accidents with Agricultural Machinery in Austria.

    PubMed

    Kogler, Robert; Quendler, Elisabeth; Boxberger, Josef

    2016-01-01

    The number of recognized accidents with fatalities during agricultural and forestry work, despite better technology and coordinated prevention and trainings, is still very high in Austria. The accident scenarios in which people are injured are very different on farms. The common causes of accidents in agriculture and forestry are the loss of control of machine, means of transport or handling equipment, hand-held tool, and object or animal, followed by slipping, stumbling and falling, breakage, bursting, splitting, slipping, fall, and collapse of material agent. In the literature, a number of studies of general (machine- and animal-related accidents) and specific (machine-related accidents) agricultural and forestry accident situations can be found that refer to different databases. From the database Data of the Austrian Workers Compensation Board (AUVA) about occupational accidents with different agricultural machinery over the period 2008-2010 in Austria, main characteristics of the accident, the victim, and the employer as well as variables on causes and circumstances by frequency and contexts of parameters were statistically analyzed by employing the chi-square test and odds ratio. The aim of the study was to determine the information content and quality of the European Statistics on Accidents at Work (ESAW) variables to evaluate safety gaps and risks as well as the accidental man-machine interaction. PMID:26479843

  1. Core phenomenology. TEC report on CRBRP PRA Phase II, Task 6C. Final draft report, Revision 1

    SciTech Connect

    1984-04-04

    As part of the determination of the risk potential associated with core-damage accident sequences for the CRBRP, a review of the core-damage phenomenology is necessary. How core damage proceeds, its effects on the primary system boundary, and the timing and energetic potential associated with core damage are important to determining the challenge to containment and the ultimate release of fission products to the environment. This chapter addresses the phenomenology related to the core-damage processes and by the use of a core-response event tree, estimates are made of the probability that certain core-response scenarios are followed.

  2. Core-core and core-valence correlation

    NASA Technical Reports Server (NTRS)

    Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

    1988-01-01

    The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

  3. 41 CFR 101-39.407 - Accident records.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 41 Public Contracts and Property Management 2 2011-07-01 2007-07-01 true Accident records. 101-39...-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.4-Accidents and Claims § 101-39.407 Accident records. If GSA's records of vehicle accidents indicate that a particular activity has had an unusually high accident...

  4. 22 CFR 102.17 - Reports on accident.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 22 Foreign Relations 1 2011-04-01 2011-04-01 false Reports on accident. 102.17 Section 102.17... Accidents Abroad Foreign Aircraft Accidents Involving United States Persons Or Property § 102.17 Reports on accident. When an accident occurs to a foreign aircraft in the district of a Foreign Service post...

  5. 22 CFR 102.17 - Reports on accident.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Reports on accident. 102.17 Section 102.17... Accidents Abroad Foreign Aircraft Accidents Involving United States Persons Or Property § 102.17 Reports on accident. When an accident occurs to a foreign aircraft in the district of a Foreign Service post...

  6. 41 CFR 101-39.407 - Accident records.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Accident records. 101-39...-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.4-Accidents and Claims § 101-39.407 Accident records. If GSA's records of vehicle accidents indicate that a particular activity has had an unusually high accident...

  7. The Accident at Fukushima: What Happened?

    SciTech Connect

    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

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

    SciTech Connect

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

    2013-09-01

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

  9. Analysis of Kuosheng Large-Break Loss-of-Coolant Accident with MELCOR 1.8.4

    SciTech Connect

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

    2000-09-15

    The MELCOR code, developed by Sandia National Laboratories, is capable of simulating the severe accident phenomena of light water reactor nuclear power plants (NPPs). A specific large-break loss-of-coolant accident (LOCA) for Kuosheng NPP is simulated with the use of the MELCOR 1.8.4 code. This accident is induced by a double-ended guillotine break of one of the recirculation pipes concurrent with complete failure of the emergency core cooling system. The MELCOR input deck for the Kuosheng NPP is established based on the design data of the Kuosheng NPP and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The effect of the MELCOR 1.8.4-provided initialization process is demonstrated. The main severe accident phenomena and the corresponding fission product released fractions associated with the large-break LOCA sequences are simulated. The MELCOR 1.8.4 predicts a longer time interval between the core collapse and vessel failure and a higher source term. This MELCOR 1.8.4 input deck will be applied to the probabilistic risk assessment, the severe accident analysis, and the severe accident management study of the Kuosheng NPP in the near future.

  10. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Accident notification. 659.33 Section 659.33 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL TRANSIT ADMINISTRATION... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency...

  11. A Serious Game for Traffic Accident Investigators

    ERIC Educational Resources Information Center

    Binsubaih, Ahmed; Maddock, Steve; Romano, Daniela

    2006-01-01

    In Dubai, traffic accidents kill one person every 37 hours and injure one person every 3 hours. Novice traffic accident investigators in the Dubai police force are expected to "learn by doing" in this intense environment. Currently, they use no alternative to the real world in order to practice. This paper argues for the use of an alternative…

  12. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration representatives. If a scheduled United States air carrier is involved the airline representatives concerned will... promptly to the nearest office of the airline concerned and to the nearest office of the Civil...

  13. Traffic accidents, facial injuries, and psychiatry.

    PubMed

    Berger, J C

    1975-01-01

    The extent to which emotional factors play a direct or indirect role in the causation of traffic accidents has been presented along with the early and late emotional response of individuals to facial injuries as a result of traffic accidents. Illustrated case histories are presented. PMID:1116323

  14. 48 CFR 36.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Accident prevention. 36.513 Section 36.513 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION SPECIAL... prevention. (a) The contracting officer shall insert the clause at 52.236-13, Accident Prevention,...

  15. 76 FR 55079 - Recreational Vessel Accident Reporting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-06

    ...The Coast Guard has received recommendations from the National Boating Safety Advisory Council (NBSAC) regarding potential ways to improve the recreational boating accident reporting process. NBSAC recommended that the Coast Guard: (1) Use a two-tiered reporting system for boating accidents; and (2) take steps to clarify what, how, and when information is reported. This notice solicits public......

  16. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Accident notification. 659.33 Section 659.33 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL TRANSIT ADMINISTRATION... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency...

  17. Normal Accident at Three Mile Island.

    ERIC Educational Resources Information Center

    Perrow, Charles

    1981-01-01

    Discusses some aspects of the accident at the Three Mile Island nuclear power plant. Explains a number of factors involved including the type of accident, warnings, design and equipment failure, operator error, and negative synergy. Presents alternatives to systems with catastrophic potential. (MK)

  18. Commercial SNF Accident Release Fractions

    SciTech Connect

    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

  19. Road accidents caused by drivers falling asleep.

    PubMed

    Sagberg, F

    1999-11-01

    About 29600 Norwegian accident-involved drivers received a questionnaire about the last accident reported to their insurance company. About 9200 drivers (31%) returned the questionnaire. The questionnaire contained questions about sleep or fatigue as contributing factors to the accident. In addition, the drivers reported whether or not they had fallen asleep some time whilst driving. and what the consequences had been. Sleep or drowsiness was a contributing factor in 3.9% of all accidents, as reported by drivers who were at fault for the accident. This factor was strongly over-represented in night-time accidents (18.6%), in running-off-the-road accidents (8.3%), accidents after driving more than 150 km on one trip (8.1%), and personal injury accidents (7.3%). A logistic regression analysis showed that the following additional factors made significant and independent contributions to increasing the odds of sleep involvement in an accident: dry road, high speed limit, driving one's own car, not driving the car daily, high education, and few years of driving experience. More male than female drivers were involved in sleep-related accidents, but this seems largely to be explained by males driving relatively more than females on roads with high speed limits. A total of 10% of male drivers and 4% of females reported to have fallen asleep while driving during the last 12 months. A total of 4% of these events resulted in an accident. The most frequent consequence of falling asleep--amounting to more than 40% of the reported incidents--was crossing of the right edge-line before awaking, whereas crossing of the centreline was reported by 16%. Drivers' lack of awareness of important precursors of falling asleep--like highway hypnosis, driving without awareness, and similar phenomena--as well as a reluctance to discontinue driving despite feeling tired are pointed out as likely contributors to sleep-related accidents. More knowledge about the drivers' experiences immediately

  20. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect

    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

  1. Human Factors in Cabin Accident Investigations

    NASA Technical Reports Server (NTRS)

    Chute, Rebecca D.; Rosekind, Mark R. (Technical Monitor)

    1996-01-01

    Human factors has become an integral part of the accident investigation protocol. However, much of the investigative process remains focussed on the flight deck, airframe, and power plant systems. As a consequence, little data has been collected regarding the human factors issues within and involving the cabin during an accident. Therefore, the possibility exists that contributing factors that lie within that domain may be overlooked. The FAA Office of Accident Investigation is sponsoring a two-day workshop on cabin safety accident investigation. This course, within the workshop, will be of two hours duration and will explore relevant areas of human factors research. Specifically, the three areas of discussion are: Information transfer and resource management, fatigue and other physical stressors, and the human/machine interface. Integration of these areas will be accomplished by providing a suggested checklist of specific cabin-related human factors questions for investigators to probe following an accident.

  2. Ethnic differences in accident rates at work.

    PubMed Central

    Baker, C C

    1987-01-01

    The accidents at work of 4482 employees in a car engine machining and assembly plant in south east England were studied retrospectively over a 12 month period. The study population was composed of Asian (22%), white (66%), and West Indian employees (12%). The crude accident rates differed among the groups, the means being Asians 1.58, white 1.23, and West Indians 1.28. There was, however, no consistent ethnic difference after adjustment for other factors such as age, type of job, and duration of service. Accident rates were higher in those employees who were younger, newly employed, and in production jobs. The findings of this research imply that accident prevention programmes should be directed to those factors known to relate to accidents and not to any specific ethnic group. PMID:3828246

  3. Pilot-error accidents: male vs female.

    PubMed

    Vail, G J; Ekman, L G

    1986-12-01

    In this study, general aviation accident records from the files of the National Transportation Safety Board (NTSB), have been analysed by gender to observe the number and rate of pilot-error related accidents from 1972 to 1981 inclusive. If both females and males have no difference in performance, then data would have indicated similarities of accident rates and types of injuries. Males had a higher rate of accidents than females, and a higher portion of the male accidents resulted in fatalities or serious injuries than for females. Type of certificate, age, total flight time, flight time in type of aircraft, phase of operation, category of flying, degree of injury, specific cause factors, cause factor miscellaneous acts/conditions were analysed, taking the total number of United States Active Civilian General Aviation Pilots into consideration. The data did indicate a difference in all variables. PMID:15676598

  4. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect

    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.

  5. The modeling of core melting and in-vessel corium relocation in the APRIL code

    SciTech Connect

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T.

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  6. European Pressurized water Reactor (EPR) SAR ATWS Accident Analyses by using 3D Code Internal Coupling Method

    SciTech Connect

    Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.

    2004-07-01

    Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)

  7. National registration of accidents in Iceland.

    PubMed

    Olafsson, O; Axelsson, J

    1992-01-01

    Community based registration of accidents has been employed in Iceland from 1987. A form developed in the emergency ward at the city Hospital of Reykjavik has been used for the registration. The following issues have been registered: the type and the seriousness of the injury, treatment, place of accident and time of accident. Health centres in Iceland have been computerized from 1976. At the time being about half of the health centres participate in the registration with the information included in the form as the source. Every health center has its well defined district. The accidents among the inhabitants in each district is registered, while accidents among other people, e.g. tourists, is registered separately. At this moment 183,000 out of a total number of 259,000 inhabitants are covered by the registration, i.e. 71% of the population. In 1989 the frequency of accidents was 198 per 100,000 inhabitants. 26% of the accidents occurred at home, 11% at work, 9% during physical activity, 6% was traffic accidents, whereas the same proportion occurred at school. This registration system has been created as a result of annual conferences on accidents arranged by the Director General of public health since 1984. Representatives for the following parties have been invited; medical doctors working in hospitals and health centres, clinical nurses, physiotherapists, the National Insurance Service, other insurance companies, rescue and ambulance personal, fire departments, the Automobile Association, the communication Council. Local communities members of the parliament, voluntary organizations, e.g. Red Cross, the Sea Rescue Service and the Aviation Board. This activity has stimulated measures aiming at preventing accidents in several local communities.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:1285816

  8. Pre-conceptual design study of ASTRID core

    SciTech Connect

    Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L.; Scholer, A. C.; Verrier, D.

    2012-07-01

    In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

  9. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  10. Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

    SciTech Connect

    Pint, Bruce A.; Unocic, Kinga A.; Terrani, Kurt A.

    2015-08-01

    Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 [1-3] and this facility continues to support those efforts in the ATF community. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [4-5]. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.

  11. Academic Rigor: The Core of the Core

    ERIC Educational Resources Information Center

    Brunner, Judy

    2013-01-01

    Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

  12. The Fukushima accident was preventable.

    PubMed

    Synolakis, Costas; Kânoğlu, Utku

    2015-10-28

    The 11 March 2011 tsunami was probably the fourth largest in the past 100 years and killed over 15 000 people. The magnitude of the design tsunami triggering earthquake affecting this region of Japan had been grossly underestimated, and the tsunami hit the Fukushima Dai-ichi nuclear power plant (NPP), causing the third most severe accident in an NPP ever. Interestingly, while the Onagawa NPP was also hit by a tsunami of approximately the same height as Dai-ichi, it survived the event 'remarkably undamaged'. We explain what has been referred to as the cascade of engineering and regulatory failures that led to the Fukushima disaster. One, insufficient attention had been given to evidence of large tsunamis inundating the region earlier, to Japanese research suggestive that large earthquakes could occur anywhere along a subduction zone, and to new research on mega-thrusts since Boxing Day 2004. Two, there were unexplainably different design conditions for NPPs at close distances from each other. Three, the hazard analysis to calculate the maximum probable tsunami at Dai-ichi appeared to have had methodological mistakes, which almost nobody experienced in tsunami engineering would have made. Four, there were substantial inadequacies in the Japan nuclear regulatory structure. The Fukushima accident was preventable, if international best practices and standards had been followed, if there had been international reviews, and had common sense prevailed in the interpretation of pre-existing geological and hydrodynamic findings. Formal standards are needed for evaluating the tsunami vulnerability of NPPs, for specific training of engineers and scientists who perform tsunami computations for emergency preparedness or critical facilities, as well as for regulators who review safety studies. PMID:26392611

  13. Assessment of severe accident prevention and mitigation features: BWR (boiling water reactor), Mark I containment design

    SciTech Connect

    Pratt, W.T.; Eltawila, F.; Perkins, K.R.; Fitzpatrick, R.G.; Luckas, W.J.; Lehner, J.R.; Davis, P.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark I containments (BWR Mark I's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Peach Bottom plant and from assessment of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark I to severe accident containment loads were also identified. In addition, those features of a BWR Mark I, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Peach Bottom and other Mark I plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance.

  14. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema

    None

    2014-03-11

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  15. SAMPSON Parallel Computation for Sensitivity Analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant Accident

    NASA Astrophysics Data System (ADS)

    Pellegrini, M.; Bautista Gomez, L.; Maruyama, N.; Naitoh, M.; Matsuoka, S.; Cappello, F.

    2014-06-01

    On March 11th 2011 a high magnitude earthquake and consequent tsunami struck the east coast of Japan, resulting in a nuclear accident unprecedented in time and extents. After scram started at all power stations affected by the earthquake, diesel generators began operation as designed until tsunami waves reached the power plants located on the east coast. This had a catastrophic impact on the availability of plant safety systems at TEPCO's Fukushima Daiichi, leading to the condition of station black-out from unit 1 to 3. In this article the accident scenario is studied with the SAMPSON code. SAMPSON is a severe accident computer code composed of hierarchical modules to account for the diverse physics involved in the various phases of the accident evolution. A preliminary parallelization analysis of the code was performed using state-of-the-art tools and we demonstrate how this work can be beneficial to the nuclear safety analysis. This paper shows that inter-module parallelization can reduce the time to solution by more than 20%. Furthermore, the parallel code was applied to a sensitivity study for the alternative water injection into TEPCO's Fukushima Daiichi unit 3. Results show that the core melting progression is extremely sensitive to the amount and timing of water injection, resulting in a high probability of partial core melting for unit 3.

  16. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    SciTech Connect

    2013-09-25

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    2012-11-01

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

  19. Upper respiratory tract illnesses and accidents.

    PubMed

    Smith, A P; Harvey, I; Richmond, P; Peters, T J; Thomas, M; Brockman, P

    1994-07-01

    Anecdotal accounts suggest that colds and influenza may increase human error. This view is supported by laboratory studies of the effects of upper respiratory tract illnesses (URTIs) on performance efficiency, which have shown that both experimentally induced and naturally occurring URTIs reduce aspects of performance efficiency. The present research examined the relationship between accidents and URTIs by studying 923 patients attending an Accident and Emergency department at a time of year when upper respiratory tract viruses were circulating. The results revealed no significant associations between URTIs and workplace accidents, and, similarly, no significant associations emerged when all accidents were compared with other attenders. The only effect which was close to statistical significance was a protective effect of influenza against workplace accidents, which could be explained in terms of a person with influenza or who has recently had influenza being less likely to work and therefore less likely to be at risk of experiencing a workplace accident. Further research must examine this topic with different methodologies, such as selecting controls from fellow workers of the index case, and these studies will provide us with a clearer view as to whether or not there is an association between URTIs and workplace accidents. PMID:7919298

  20. An analysis of aircraft accidents involving fires

    NASA Technical Reports Server (NTRS)

    Lucha, G. V.; Robertson, M. A.; Schooley, F. A.

    1975-01-01

    All U. S. Air Carrier accidents between 1963 and 1974 were studied to assess the extent of total personnel and aircraft damage which occurred in accidents and in accidents involving fire. Published accident reports and NTSB investigators' factual backup files were the primary sources of data. Although it was frequently not possible to assess the relative extent of fire-caused damage versus impact damage using the available data, the study established upper and lower bounds for deaths and damage due specifically to fire. In 12 years there were 122 accidents which involved airframe fires. Eighty-seven percent of the fires occurred after impact, and fuel leakage from ruptured tanks or severed lines was the most frequently cited cause. A cost analysis was performed for 300 serious accidents, including 92 serious accidents which involved fire. Personal injury costs were outside the scope of the cost analysis, but data on personnel injury judgements as well as settlements received from the CAB are included for reference.

  1. Accidents in Canada: mortality and hospitalization.

    PubMed

    Riley, R; Paddon, P

    1989-01-01

    For Canadians under 45, accidents are the leading cause of both death and hospitalization. For the Canadian population as a whole, accidents rank fourth as a cause of death, after cardiovascular disease (CVD), cancer and respiratory disease. This article analyzes accident mortality and hospitalization in Canada using age-specific rates, age-standardized mortality rates (ASMR), and potential years of life lost (PYLL). The six major causes of accidental death for men are motor vehicle traffic accidents (MVTA), falls, drowning, fires, suffocation and poisoning. For women, the order is slightly different: MVTA, falls, fires, suffocation, poisoning and drowning. From 1971 to 1986, age-standardized mortality rates (ASMR) for accidents decreased by 44% for men and 39% for women. The largest decrease occurred in the under 15 age group. Accidents accounted for 11.5% of total hospital days in 1985, and 8% of hospital discharges. Because young people have the highest rates of accidental death, potential years of life lost (PYLL) are almost as high for accidents as for cardiovascular disease, although CVD deaths outnumbered accidental deaths by almost five to one in 1985. PMID:2491351

  2. Markov Model of Accident Progression at Fukushima Daiichi

    SciTech Connect

    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.

  3. Analysis of Radionuclide Releases from the Fukushima Dai-ichi Nuclear Power Plant Accident Part II

    NASA Astrophysics Data System (ADS)

    Achim, Pascal; Monfort, Marguerite; Le Petit, Gilbert; Gross, Philippe; Douysset, Guilhem; Taffary, Thomas; Blanchard, Xavier; Moulin, Christophe

    2014-03-01

    The present part of the publication (Part II) deals with long range dispersion of radionuclides emitted into the atmosphere during the Fukushima Dai-ichi accident that occurred after the March 11, 2011 tsunami. The first part (Part I) is dedicated to the accident features relying on radionuclide detections performed by monitoring stations of the Comprehensive Nuclear Test Ban Treaty Organization network. In this study, the emissions of the three fission products Cs-137, I-131 and Xe-133 are investigated. Regarding Xe-133, the total release is estimated to be of the order of 6 × 1018 Bq emitted during the explosions of units 1, 2 and 3. The total source term estimated gives a fraction of core inventory of about 8 × 1018 Bq at the time of reactors shutdown. This result suggests that at least 80 % of the core inventory has been released into the atmosphere and indicates a broad meltdown of reactor cores. Total atmospheric releases of Cs-137 and I-131 aerosols are estimated to be 1016 and 1017 Bq, respectively. By neglecting gas/particulate conversion phenomena, the total release of I-131 (gas + aerosol) could be estimated to be 4 × 1017 Bq. Atmospheric transport simulations suggest that the main air emissions have occurred during the events of March 14, 2011 (UTC) and that no major release occurred after March 23. The radioactivity emitted into the atmosphere could represent 10 % of the Chernobyl accident releases for I-131 and Cs-137.

  4. Accident source terms for Light-Water Nuclear Power Plants. Final report

    SciTech Connect

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ``Calculation of Distance Factors for Power and Test Reactors`` which specified a release of fission products from the core to the reactor containment for a postulated accident involving ``substantial meltdown of the core``. This ``source term``, tile basis for tile NRC`s Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC`s reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ``source term`` release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ``source term`` is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it.

  5. Coring Sample Acquisition Tool

    NASA Technical Reports Server (NTRS)

    Haddad, Nicolas E.; Murray, Saben D.; Walkemeyer, Phillip E.; Badescu, Mircea; Sherrit, Stewart; Bao, Xiaoqi; Kriechbaum, Kristopher L.; Richardson, Megan; Klein, Kerry J.

    2012-01-01

    A sample acquisition tool (SAT) has been developed that can be used autonomously to sample drill and capture rock cores. The tool is designed to accommodate core transfer using a sample tube to the IMSAH (integrated Mars sample acquisition and handling) SHEC (sample handling, encapsulation, and containerization) without ever touching the pristine core sample in the transfer process.

  6. The Core Skills Initiative.

    ERIC Educational Resources Information Center

    Further Education Unit, London (England).

    A British initiative that aims to identify, develop, and assess core skills in post-16 courses and qualifications is summarized in this bulletin. The first section discusses expectations regarding what core skills can achieve. The following section focuses on other purposes to which core skills could contribute, such as broadening the post-16…

  7. Core Competence and Education.

    ERIC Educational Resources Information Center

    Holmes, Gary; Hooper, Nick

    2000-01-01

    Outlines the concept of core competence and applies it to postcompulsory education in the United Kingdom. Adopts an educational perspective that suggests accreditation as the core competence of universities. This economic approach suggests that the market trend toward lifetime learning might best be met by institutions developing a core competence…

  8. Core Design Applications

    Energy Science and Technology Software Center (ESTSC)

    1995-07-12

    CORD-2 is intended for core desigh applications of pressurized water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refueling).

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

    SciTech Connect

    Wenxi Tian; Suizheng Qiu; Guanghui Su; Dounan Jia; 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)

  10. An Application of CICCT Accident Categories to Aviation Accidents in 1988-2004

    NASA Technical Reports Server (NTRS)

    Evans, Joni K.

    2007-01-01

    Interventions or technologies developed to improve aviation safety often focus on specific causes or accident categories. Evaluation of the potential effectiveness of those interventions is dependent upon mapping the historical aviation accidents into those same accident categories. To that end, the United States civil aviation accidents occurring between 1988 and 2004 (n=26,117) were assigned accident categories based upon the taxonomy developed by the CAST/ICAO Common Taxonomy Team (CICTT). Results are presented separately for four main categories of flight rules: Part 121 (large commercial air carriers), Scheduled Part 135 (commuter airlines), Non-Scheduled Part 135 (on-demand air taxi) and Part 91 (general aviation). Injuries and aircraft damage are summarized by year and by accident category.

  11. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    SciTech Connect

    Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior, Antonio Jr.; Silva Rocha, Marcelo da; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; Souza Lima, Ana Cecilia de

    2013-05-06

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm{sup 2}, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  12. Shuttle accident stalls science plans

    NASA Astrophysics Data System (ADS)

    Katzoff, Judith A.

    Plans to make 1986 a uniquely productive year for U.S. space science activities ended in one horrible moment with the January 28, 1986, explosion of the space shuttle Challenger. The joyless scene at Cape Canaveral, Fla., stood in sharp contrast to the overwhelming success of Voyager 2 in its encounter with Uranus 4 days earlier. (Scientific details of that encounter will follow in upcoming issues of Eos.)Of the 15 space shuttle flights planned for fiscal year 1986, beginning October 1, 1985, a total of seven were to have carried scientific payloads for the National Aeronautics and Space Administration (NASA). The remaining eight flights were evenly divided between missions for the U.S. Department of Defense and commercial missions for NASA's paying customers. The explosion caused NASA to put its entire space shuttle program on hold to allow time for engineers to find the cause of the accident and for NASA to implement corrective measures. As Eos went to press, NASA acting administrator William R. Graham had not yet released the names of those who would serve on the formal investigative panel. “I think everybody's agreed that it will take weeks to months to unravel,” said Alexander Dessler, director of the space science laboratory at NASA's Marshall Space Flight Center near Huntsville, Ala. Dessler speculated that investigators would begin with a list of hundreds of possible causes for the explosion.

  13. Banded transformer cores

    NASA Technical Reports Server (NTRS)

    Mclyman, C. W. T. (Inventor)

    1974-01-01

    A banded transformer core formed by positioning a pair of mated, similar core halves on a supporting pedestal. The core halves are encircled with a strap, selectively applying tension whereby a compressive force is applied to the core edge for reducing the innate air gap. A dc magnetic field is employed in supporting the core halves during initial phases of the banding operation, while an ac magnetic field subsequently is employed for detecting dimension changes occurring in the air gaps as tension is applied to the strap.

  14. Safety analysis of surface haulage accidents

    SciTech Connect

    Randolph, R.F.; Boldt, C.M.K.

    1996-12-31

    Research on improving haulage truck safety, started by the U.S. Bureau of Mines, is being continued by its successors. This paper reports the orientation of the renewed research efforts, beginning with an update on accident data analysis, the role of multiple causes in these accidents, and the search for practical methods for addressing the most important causes. Fatal haulage accidents most often involve loss of control or collisions caused by a variety of factors. Lost-time injuries most often involve sprains or strains to the back or multiple body areas, which can often be attributed to rough roads and the shocks of loading and unloading. Research to reduce these accidents includes improved warning systems, shock isolation for drivers, encouraging seatbelt usage, and general improvements to system and task design.

  15. Review of models applicable to accident aerosols

    SciTech Connect

    Glissmeyer, J.A.

    1983-07-01

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

  16. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-01-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  17. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-11-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  18. 48 CFR 636.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... has any questions regarding construction safety issues. ... CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 636.513 Accident prevention. (a) In... when awarding construction contracts. The contracting officer shall confer with OBO/OM/SHEM if...

  19. More Children Accidently Poisoned by 'Essential Oils'

    MedlinePlus

    ... 158837.html More Children Accidently Poisoned by 'Essential Oils' Tennessee poison center reports doubling of dangerous exposures ... News) -- Children are increasingly at risk from essential oils that are often used in natural remedies, a ...

  20. Modeling secondary accidents identified by traffic shock waves.

    PubMed

    Junhua, Wang; Boya, Liu; Lanfang, Zhang; Ragland, David R

    2016-02-01

    The high potential for occurrence and the negative consequences of secondary accidents make them an issue of great concern affecting freeway safety. Using accident records from a three-year period together with California interstate freeway loop data, a dynamic method for more accurate classification based on the traffic shock wave detecting method was used to identify secondary accidents. Spatio-temporal gaps between the primary and secondary accident were proven be fit via a mixture of Weibull and normal distribution. A logistic regression model was developed to investigate major factors contributing to secondary accident occurrence. Traffic shock wave speed and volume at the occurrence of a primary accident were explicitly considered in the model, as a secondary accident is defined as an accident that occurs within the spatio-temporal impact scope of the primary accident. Results show that the shock waves originating in the wake of a primary accident have a more significant impact on the likelihood of a secondary accident occurrence than the effects of traffic volume. Primary accidents with long durations can significantly increase the possibility of secondary accidents. Unsafe speed and weather are other factors contributing to secondary crash occurrence. It is strongly suggested that when police or rescue personnel arrive at the scene of an accident, they should not suddenly block, decrease, or unblock the traffic flow, but instead endeavor to control traffic in a smooth and controlled manner. Also it is important to reduce accident processing time to reduce the risk of secondary accident. PMID:26687540

  1. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 1 2012-10-01 2012-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident means— (a) Any casualty or accident involving...

  2. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 1 2014-10-01 2014-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident means— (a) Any casualty or accident involving...

  3. 32 CFR 634.29 - Traffic accident investigation reports.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 4 2012-07-01 2011-07-01 true Traffic accident investigation reports. 634.29... accident investigation reports. (a) Accidents requiring immediate reports. The driver or owner of any vehicle involved in an accident, as described in § 634.28, on the installation, must immediately...

  4. 32 CFR 634.29 - Traffic accident investigation reports.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 4 2011-07-01 2011-07-01 false Traffic accident investigation reports. 634.29... accident investigation reports. (a) Accidents requiring immediate reports. The driver or owner of any vehicle involved in an accident, as described in § 634.28, on the installation, must immediately...

  5. 32 CFR 634.29 - Traffic accident investigation reports.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 4 2014-07-01 2013-07-01 true Traffic accident investigation reports. 634.29... accident investigation reports. (a) Accidents requiring immediate reports. The driver or owner of any vehicle involved in an accident, as described in § 634.28, on the installation, must immediately...

  6. 32 CFR 634.29 - Traffic accident investigation reports.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 4 2013-07-01 2013-07-01 false Traffic accident investigation reports. 634.29... accident investigation reports. (a) Accidents requiring immediate reports. The driver or owner of any vehicle involved in an accident, as described in § 634.28, on the installation, must immediately...

  7. 33 CFR 401.81 - Reporting an accident.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 3 2011-07-01 2011-07-01 false Reporting an accident. 401.81... an accident. (a) Where a vessel on the Seaway is involved in an accident or a dangerous occurrence, the master of the vessel shall report the accident or occurrence, pursuant to the requirements of...

  8. 48 CFR 852.236-87 - Accident prevention.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Accident prevention. 852... Accident prevention. As prescribed in 836.513, insert the following clause: Accident Prevention (SEP 1993....236-13, Accident Prevention. However, only the Contracting Officer may issue an order to stop all...

  9. 14 CFR 415.41 - Accident investigation plan.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 14 Aeronautics and Space 4 2011-01-01 2011-01-01 false Accident investigation plan. 415.41 Section... Launch Range § 415.41 Accident investigation plan. An applicant must file an accident investigation plan... reporting and responding to launch accidents, launch incidents, or other mishaps, as defined by § 401.5...

  10. 46 CFR 97.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Accidents to machinery. 97.30-5 Section 97.30-5 Shipping... Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  11. 41 CFR 101-39.401 - Reporting of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 41 Public Contracts and Property Management 2 2011-07-01 2007-07-01 true Reporting of accidents...-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.4-Accidents and Claims § 101-39.401 Reporting of accidents. (a) The..., by telephone, or by facsimile machine of any accident in which the vehicle may be involved: (1)...

  12. 32 CFR 634.28 - Traffic accident investigation.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 4 2011-07-01 2011-07-01 false Traffic accident investigation. 634.28 Section... accident investigation. Installation law enforcement personnel must make detailed investigations of accidents described in this section: (a) Accidents involving Government vehicles or Government property...

  13. 49 CFR 382.209 - Use following an accident.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 5 2011-10-01 2011-10-01 false Use following an accident. 382.209 Section 382.209... ALCOHOL USE AND TESTING Prohibitions § 382.209 Use following an accident. No driver required to take a post-accident alcohol test under § 382.303 shall use alcohol for eight hours following the accident,...

  14. 46 CFR 196.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Accidents to machinery. 196.30-5 Section 196.30-5... Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  15. 49 CFR 801.30 - Records from accident investigations.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Records from accident investigations. 801.30... TRANSPORTATION SAFETY BOARD PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.30 Records from accident investigations. Upon completion of an accident investigation, each NTSB investigator...

  16. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager... 50 Wildlife and Fisheries 8 2011-10-01 2011-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to...

  17. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to...

  18. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  19. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 4 2011-01-01 2011-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  20. 48 CFR 852.236-87 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Accident prevention. 852... Accident prevention. As prescribed in 836.513, insert the following clause: Accident Prevention (SEP 1993....236-13, Accident Prevention. However, only the Contracting Officer may issue an order to stop all...

  1. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 16 2012-07-01 2012-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  2. 49 CFR 382.209 - Use following an accident.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 5 2010-10-01 2010-10-01 false Use following an accident. 382.209 Section 382.209... ALCOHOL USE AND TESTING Prohibitions § 382.209 Use following an accident. No driver required to take a post-accident alcohol test under § 382.303 shall use alcohol for eight hours following the accident,...

  3. 33 CFR 401.81 - Reporting an accident.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false Reporting an accident. 401.81... an accident. (a) Where a vessel on the Seaway is involved in an accident or a dangerous occurrence, the master of the vessel shall report the accident or occurrence, pursuant to the requirements of...

  4. 33 CFR 173.55 - Report of casualty or accident.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Report of casualty or accident... (CONTINUED) BOATING SAFETY VESSEL NUMBERING AND CASUALTY AND ACCIDENT REPORTING Casualty and Accident Reporting § 173.55 Report of casualty or accident. (a) The operator of a vessel shall submit the casualty...

  5. 49 CFR 801.30 - Records from accident investigations.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Records from accident investigations. 801.30... TRANSPORTATION SAFETY BOARD PUBLIC AVAILABILITY OF INFORMATION Accident Investigation Records § 801.30 Records from accident investigations. Upon completion of an accident investigation, each NTSB investigator...

  6. 14 CFR 415.41 - Accident investigation plan.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Accident investigation plan. 415.41 Section... Launch Range § 415.41 Accident investigation plan. An applicant must file an accident investigation plan... reporting and responding to launch accidents, launch incidents, or other mishaps, as defined by § 401.5...

  7. 32 CFR 634.28 - Traffic accident investigation.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 4 2010-07-01 2010-07-01 true Traffic accident investigation. 634.28 Section... accident investigation. Installation law enforcement personnel must make detailed investigations of accidents described in this section: (a) Accidents involving Government vehicles or Government property...

  8. 41 CFR 101-39.401 - Reporting of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Reporting of accidents...-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.4-Accidents and Claims § 101-39.401 Reporting of accidents. (a) The..., by telephone, or by facsimile machine of any accident in which the vehicle may be involved: (1)...

  9. Oranges and Peaches: Understanding Communication Accidents in the Reference Interview.

    ERIC Educational Resources Information Center

    Dewdney, Patricia; Michell, Gillian

    1996-01-01

    Librarians often have communication "accidents" with reference questions as initially presented. This article presents linguistic analysis of query categories, including: simple failures of hearing, accidents involving pronunciation or homophones, accidents where users repeat earlier misinterpretations to librarians, and accidents where users…

  10. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 15 2011-07-01 2011-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  11. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  12. 49 CFR 655.44 - Post-accident testing.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Safety Administration rule 49 CFR 389.303(a)(1) or (b)(1). (ii) The employer shall also drug and alcohol... 49 Transportation 7 2012-10-01 2012-10-01 false Post-accident testing. 655.44 Section 655.44... of Testing § 655.44 Post-accident testing. (a) Accidents. (1) Fatal accidents. (i) As soon...

  13. 49 CFR 655.44 - Post-accident testing.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Safety Administration rule 49 CFR 389.303(a)(1) or (b)(1). (ii) The employer shall also drug and alcohol... 49 Transportation 7 2011-10-01 2011-10-01 false Post-accident testing. 655.44 Section 655.44... of Testing § 655.44 Post-accident testing. (a) Accidents. (1) Fatal accidents. (i) As soon...

  14. 49 CFR 655.44 - Post-accident testing.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Safety Administration rule 49 CFR 389.303(a)(1) or (b)(1). (ii) The employer shall also drug and alcohol... 49 Transportation 7 2013-10-01 2013-10-01 false Post-accident testing. 655.44 Section 655.44... of Testing § 655.44 Post-accident testing. (a) Accidents. (1) Fatal accidents. (i) As soon...

  15. 49 CFR 655.44 - Post-accident testing.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Safety Administration rule 49 CFR 389.303(a)(1) or (b)(1). (ii) The employer shall also drug and alcohol... 49 Transportation 7 2010-10-01 2010-10-01 false Post-accident testing. 655.44 Section 655.44... of Testing § 655.44 Post-accident testing. (a) Accidents. (1) Fatal accidents. (i) As soon...

  16. 49 CFR 655.44 - Post-accident testing.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Safety Administration rule 49 CFR 389.303(a)(1) or (b)(1). (ii) The employer shall also drug and alcohol... 49 Transportation 7 2014-10-01 2014-10-01 false Post-accident testing. 655.44 Section 655.44... of Testing § 655.44 Post-accident testing. (a) Accidents. (1) Fatal accidents. (i) As soon...

  17. 46 CFR 97.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Accidents to machinery. 97.30-5 Section 97.30-5 Shipping... Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  18. 46 CFR 196.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Accidents to machinery. 196.30-5 Section 196.30-5... Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use...

  19. 46 CFR 78.33-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Accidents to machinery. 78.33-5 Section 78.33-5 Shipping... Accidents, Repairs, and Unsafe Equipment § 78.33-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of the...

  20. "Snowing" Core in Earth?

    NASA Astrophysics Data System (ADS)

    Li, J.; Chen, B.; Cormier, V.; Gao, L.; Gubbins, D.; Kharlamova, S. A.; He, K.; Yang, H.

    2008-12-01

    As a planet cools, an initially molten core gradually solidifies. Solidification occurs at shallow depths in the form of "snow", if the liquidus temperature gradient of the core composition is smaller than the adiabatic temperature gradient in the core. Experimental data on the melting behavior of iron-sulfur binary system suggest that the cores of Mercury and Ganymede are probably snowing at the present time. The Martian core is predicted to snow in the future, provided that the sulfur content falls into the range of 10 to 14 weight percent. Is the Earth's core snowing? If so, what are the surface manifestations? If the Earth's core snowed in the past, how did it affect the formation of the solid inner core and the geodynamo? Here, we evaluate the likelihood and consequences of a snowing core throughout the Earth's history, on the basis of mineral physics data describing the melting behavior, equation-of-state, and thermodynamic properties of iron-rich alloys at high pressures. We discuss if snowing in the present-day Earth can reproduce the shallow gradients of compressional wave velocity above the inner-core boundary, and whether or not snowing in the early Earth may reconcile the apparent young age of the solid inner core with a long-lived geodynamo.

  1. Learning lessons from Natech accidents - the eNATECH accident database

    NASA Astrophysics Data System (ADS)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

    When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

  2. Continuous coring drill bit

    SciTech Connect

    Ford, G.A.

    1987-09-22

    A continuous coring drill bit is described comprising: (a) body means defining a vertical axis and adapted for connection to drill pipe and forming an internal body cavity disposed in eccentric relation with the vertical axis and a generally circular throat in communication with the body cavity for conducting drilling fluid. The throat defining a throat axis coincident with the vertical axis and being of a configuration permitting passage of a formation core into the body cavity; (b) a generally cylindrical tubular core breaker being rotatably mounted within the body cavity and defining a vertical axis of rotation of generally parallel and offset relation with the vertical axis of the body means; and (c) a buttress element extending inwardly from the core breaker and adapted to contact the formation core. Upon each rotation of the drill bit the buttress element applying transverse force to the core for fracturing of the core into sections sufficiently small for transport by the drilling fluid.

  3. Severe accident testing of electrical penetration assemblies

    SciTech Connect

    Clauss, D.B. )

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  4. Delay in ambulance dispatch to road accidents.

    PubMed

    Brodsky, H

    1992-06-01

    When a road accident occurs, the police communications officer, or 911 operator, generally receives the first call. If the caller reports injuries, the emergency medical services dispatcher is notified immediately; but if the caller is uncertain of injuries, the operator may wait. Most often an ambulance is not needed. However, in nearly 20% of fatal road accidents in Missouri, waiting for confirmation of need resulted in a delay of 5 minutes or more in the dispatch of an ambulance. PMID:1585968

  5. Delay in ambulance dispatch to road accidents.

    PubMed Central

    Brodsky, H

    1992-01-01

    When a road accident occurs, the police communications officer, or 911 operator, generally receives the first call. If the caller reports injuries, the emergency medical services dispatcher is notified immediately; but if the caller is uncertain of injuries, the operator may wait. Most often an ambulance is not needed. However, in nearly 20% of fatal road accidents in Missouri, waiting for confirmation of need resulted in a delay of 5 minutes or more in the dispatch of an ambulance. PMID:1585968

  6. MELCOR analyses for accident progression issues

    SciTech Connect

    Dingman, S.E.; Shaffer, C.J.; Payne, A.C.; Carmel, M.K. )

    1991-01-01

    Results of calculations performed with MELCOR and HECTR in support of the NUREG-1150 study are presented in this report. The analyses examined a wide range of issues. The analyses included integral calculations covering an entire accident sequence, as well as calculations that addressed specific issues that could affect several accident sequences. The results of the analyses for Grand Gulf, Peach Bottom, LaSalle, and Sequoyah are described, and the major conclusions are summarized. 23 refs., 69 figs., 8 tabs.

  7. COMSORS: A light water reactor chemical core catcher

    SciTech Connect

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1997-02-24

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

  8. Process for controlling accidents in chemical laboratories

    SciTech Connect

    Delvin, W.L.

    1980-10-01

    Most laboratory safety programs include inspections to identify hazards and thereby control accidents. There are certain elements that must be a part of a successful inspection and control process. These are a systematic and consistent inspection procedure, a reliable evaluation of identified hazards, and effective follow-up actions. Laboratory management, through its responsibility for the total system, has a key role in the inspection and control process for follow-up actions and accepting risks. If any of the above requirements are missing, the process will be less than adequate. Understanding the relationship between accidents, hazards, and risks is important in establishing an effective inspection and control program. Hazards are potential sources of accidents (accidents waiting to happen). Associated with each is a risk, which has two components: probability and consequence. Probability refers to the likelihood that a hazard will turn into an accident and consequence is the result of such an accident. In assessing the seriousness of a hazard, both probability and consequence must be considered in terms of risk level and acceptability. This paper presents a process that can be used by laboratory management to establish an effective inspection and control program for the laboratory. A discussion of safety concepts and their relationships that affect the process is included.

  9. Anthropotechnological analysis of industrial accidents in Brazil.

    PubMed Central

    Binder, M. C.; de Almeida, I. M.; Monteau, M.

    1999-01-01

    The Brazilian Ministry of Labour has been attempting to modify the norms used to analyse industrial accidents in the country. For this purpose, in 1994 it tried to make compulsory use of the causal tree approach to accident analysis, an approach developed in France during the 1970s, without having previously determined whether it is suitable for use under the industrial safety conditions that prevail in most Brazilian firms. In addition, opposition from Brazilian employers has blocked the proposed changes to the norms. The present study employed anthropotechnology to analyse experimental application of the causal tree method to work-related accidents in industrial firms in the region of Botucatu, São Paulo. Three work-related accidents were examined in three industrial firms representative of local, national and multinational companies. On the basis of the accidents analysed in this study, the rationale for the use of the causal tree method in Brazil can be summarized for each type of firm as follows: the method is redundant if there is a predominance of the type of risk whose elimination or neutralization requires adoption of conventional industrial safety measures (firm representative of local enterprises); the method is worth while if the company's specific technical risks have already largely been eliminated (firm representative of national enterprises); and the method is particularly appropriate if the firm has a good safety record and the causes of accidents are primarily related to industrial organization and management (multinational enterprise). PMID:10680249

  10. HYDRATE CORE DRILLING TESTS

    SciTech Connect

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate formation comprised of coarse, large

  11. Thermalhydraulic processes in the reactor coolant system of a BWR (boiling water reactor) under severe accident conditions

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MSIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a brief discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermalhydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are briefly described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is presented. 12 refs., 4 figs., 1 tab.

  12. Core damage severity evaluation for pressurized water reactors by artificial intelligence methods

    NASA Astrophysics Data System (ADS)

    Mironidis, Anastasios Pantelis

    1998-12-01

    During the course of nuclear power evolution, accidents have occurred. However, in the western world, none of them had a severe impact on the public because of the design features of nuclear plants. In nuclear reactors, barriers constitute physical obstacles to uncontrolled fission product releases. These barriers are an important factor in safety analysis. During an accident, reactor safety systems become actuated to prevent the barriers from been breached. In addition, operators are required to take specified actions, meticulously depicted in emergency response procedures. In an accident, on-the-spot knowledge regarding the condition of the core is necessary. In order to make the right decisions toward mitigating the accident severity and its consequences, we need to know the status of the core [1, 3]. However, power plant instrumentation that can provide a direct indication of the status of the core during the time when core damage is a potential outcome, does not exist. Moreover, the information from instruments may have large uncertainty of various types. Thus, a very strong potential for misinterpreting incoming information exists. This research endeavor addresses the problem of evaluating the core damage severity of a Pressurized Water Reactor during a transient or an accident. An expert system has been constructed, that incorporates knowledge and reasoning of human experts. The expert system's inference engine receives incoming plant data that originate in the plethora of core-related instruments. Its knowledge base relies on several massive, multivariate fuzzy logic rule-sets, coupled with several artificial neural networks. These mathematical models have encoded information that defines possible core states, based on correlations of parameter values. The inference process classifies the core as intact, or as experiencing clad damage and/or core melting. If the system detects a form of core damage, a quantification procedure will provide a numerical

  13. 23. CORE WORKER OPERATING A COREBLOWER THAT PNEUMATICALLY FILLED CORE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    23. CORE WORKER OPERATING A CORE-BLOWER THAT PNEUMATICALLY FILLED CORE BOXES WITH RESIGN IMPREGNATED SAND AND CREATED A CORE THAT THEN REQUIRED BAKING, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  14. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    significantly exceeded QE limits for extended time periods for the low frequency STSBO sequence evaluated in this study. It is recognized that the core damage frequency (CDF) of the sequence evaluated in this scoping effort would be considerably lower if evaluations considered new FLEX equipment being installed by industry. Nevertheless, because of uncertainties in instrumentation response when exposed to conditions beyond QE limits and alternate challenges associated with different sequences that may impact sensor performance, it is recommended that additional evaluations of instrumentation performance be completed to provide confidence that operators have access to accurate, relevant, and timely information on the status of reactor systems for a broad range of challenges associated with risk important severe accident sequences.

  15. Core-Cutoff Tool

    NASA Technical Reports Server (NTRS)

    Gheen, Darrell

    2007-01-01

    A tool makes a cut perpendicular to the cylindrical axis of a core hole at a predetermined depth to free the core at that depth. The tool does not damage the surrounding material from which the core was cut, and it operates within the core-hole kerf. Coring usually begins with use of a hole saw or a hollow cylindrical abrasive cutting tool to make an annular hole that leaves the core (sometimes called the plug ) in place. In this approach to coring as practiced heretofore, the core is removed forcibly in a manner chosen to shear the core, preferably at or near the greatest depth of the core hole. Unfortunately, such forcible removal often damages both the core and the surrounding material (see Figure 1). In an alternative prior approach, especially applicable to toxic or fragile material, a core is formed and freed by means of milling operations that generate much material waste. In contrast, the present tool eliminates the damage associated with the hole-saw approach and reduces the extent of milling operations (and, hence, reduces the waste) associated with the milling approach. The present tool (see Figure 2) includes an inner sleeve and an outer sleeve and resembles the hollow cylindrical tool used to cut the core hole. The sleeves are thin enough that this tool fits within the kerf of the core hole. The inner sleeve is attached to a shaft that, in turn, can be attached to a drill motor or handle for turning the tool. This tool also includes a cutting wire attached to the distal ends of both sleeves. The cutting wire is long enough that with sufficient relative rotation of the inner and outer sleeves, the wire can cut all the way to the center of the core. The tool is inserted in the kerf until its distal end is seated at the full depth. The inner sleeve is then turned. During turning, frictional drag on the outer core pulls the cutting wire into contact with the core. The cutting force of the wire against the core increases with the tension in the wire and

  16. Lunar core: occam's razor?

    NASA Astrophysics Data System (ADS)

    Bell, Peter M.

    Whether or not the earth's moon has a core is a much bandied question. Like many notions about the properties of the moon, ideas of a lunar core changed drastically after the Apollo studies. A review of the development of these ideas was given a scholarly treatment by S.K. Runcorn recently (Nature, 304, 589-596, 1983). In contrast, L.L. Hood, C.P. Sonett, and L.J. Srnka have questioned the concept in serious detail (Nature, 307, 661-662, 1984).Whether or not the moon actually has or has had a fluid metallic core is of great consequence for a number of geophysical theories about the solar system. Most investigators concede that the possible existence of a lunar core remains one of the major unanswered, and yet most critical, questions about the moon. A lot rides on the answer: Can a lunar-sized body have a core? Is the core metallic? How is the core related to lunar magnetism and its paleomagnetism? Is or was a lunar core related to lunar volcanism? If the moon can have a core, is planetary core formation in the solar system a simple matter of gravitationally segregating metallic fragments that were formed elsewhere? Implications of the questions are without limit. There is, perhaps, no more valid issue about the moon to explore scientifically.

  17. [Drivers of advanced age in traffic accidents].

    PubMed

    Bilban, Marjan

    2002-12-01

    The elderly are vulnerable and potentially unpredictable active participants in traffic who deserve special attention. Longer life expectancy entails a greater number of senior drivers, that is, persons with various health problems and difficulties accompanying old age. At the turn of the millennium, the share of population aged 65 or more in Slovenia was around 13%, and in 25 years it will be near as much as 19%. The share of drivers from this age group was 28% a year ago, and it is expected to reach about 54%. Numerous studies have shown that there are many differences in driving attitude between the young and the elderly. The young are by large active victims, and their main offense and cause of accident is speeding, while the elderly are more passive and their main offense is ignoring and enforcing the right of way. This paper focuses on the differences in the occurrence and type of injuries between the young and the elderly drivers, based on an analysis of all road accidents in Slovenia in the period between 1998-2000. Older people (over 65) caused only 4.7% of all road accidents (16.7% of all accidents involving pedestrians, 11.5% of all involving cyclists, 2.7% involving motorcyclists and 5% of all accidents involving car drivers). Of all accidents, 89.3% were without injuries, and the fatal outcome was registered in 0.4% accidents. Among the elderly (65-74 years of age), however, this share was 1%, and rising to 2.7% with the age 75 and above. By calculating the weight index, which discriminates between minor and severe injuries, and the fatal outcome, it was established that age groups 65-74 and > or = 75 cause three and five times greater damage, respectively than age groups from 18 to 54 years. With years, psychophysical changes lead to a drop in driving ability, which in turn increases the risk of road accidents. It is true that elderly people cause less traffic accidents (and also drive less) than the young, but when they are involved in an accident

  18. Release of Pu isotopes from the Fukushima Daiichi Nuclear Power Plant accident to the marine environment was negligible.

    PubMed

    Bu, Wenting; Fukuda, Miho; Zheng, Jian; Aono, Tatsuo; Ishimaru, Takashi; Kanda, Jota; Yang, Guosheng; Tagami, Keiko; Uchida, Shigeo; Guo, Qiuju; Yamada, Masatoshi

    2014-08-19

    Atmospheric deposition of Pu isotopes from the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident has been observed in the terrestrial environment around the FDNPP site; however, their deposition in the marine environment has not been studied. The possible contamination of Pu in the marine environment has attracted great scientific and public concern. To fully understand this possible contamination of Pu isotopes from the FDNPP accident to the marine environment, we collected marine sediment core samples within the 30 km zone around the FDNPP site in the western North Pacific about two years after the accident. Pu isotopes ((239)Pu, (240)Pu, and (241)Pu) and radiocesium isotopes ((134)Cs and (137)Cs) in the samples were determined. The high activities of radiocesium and the (134)Cs/(137)Cs activity ratios with values around 1 (decay corrected to 15 March 2011) suggested that these samples were contaminated by the FDNPP accident-released radionuclides. However, the activities of (239+240)Pu and (241)Pu were low compared with the background level before the FDNPP accident. The Pu atom ratios ((240)Pu/(239)Pu and (241)Pu/(239)Pu) suggested that global fallout and the pacific proving ground (PPG) close-in fallout are the main sources for Pu contamination in the marine sediments. As Pu isotopes are particle-reactive and they can be easily incorporated with the marine sediments, we concluded that the release of Pu isotopes from the FDNPP accident to the marine environment was negligible. PMID:25050458

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

    SciTech Connect

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

    1994-12-01

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

  20. Core sample extractor

    NASA Technical Reports Server (NTRS)

    Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark

    1989-01-01

    The problem of retrieving and storing core samples from a hole drilled on the lunar surface is addressed. The total depth of the hole in question is 50 meters with a maximum diameter of 100 millimeters. The core sample itself has a diameter of 60 millimeters and will be two meters in length. It is therefore necessary to retrieve and store 25 core samples per hole. The design utilizes a control system that will stop the mechanism at a certain depth, a cam-linkage system that will fracture the core, and a storage system that will save and catalogue the cores to be extracted. The Rod Changer and Storage Design Group will provide the necessary tooling to get into the hole as well as to the core. The mechanical design for the cam-linkage system as well as the conceptual design of the storage device are described.

  1. Drudgery, accidents and injuries in Indian agriculture.

    PubMed

    Nag, Pranab Kumar; Nag, Anjali

    2004-04-01

    The Indian farming employs 225 million workforce to cover 140 million hectares of total cultivated land. In spite of rapid farm mechanization (e.g., 149 million farm machinery), the vast resource-poor family farming has primary dependence on traditional methods (e.g., 520 million hand tools and 37 million animal-drawn implements are in operation). The work drudgery, the traumatic accidents and injuries are the major concerns to examine options for ergonomics intervention and betterment of work in crop production activities. This review summarizes human energy expenditure in crop production activities, to assess the job severity, tools and machinery, and formulate the basis to reorganize work and work methods. While the farm mechanization is more in the northern India, the accidents were more in the villages in southern India. On average of the four regions, the tractor incidents (overturning, falling from the tractor, etc.) were highest (27.7%), followed by thresher (14.6%), sprayer/duster (12.2%), sugarcane crusher (8.1%) and chaff cutter (7.8%) accidents. Most of the fatal accidents resulted from the powered machinery, with the annual fatality rate estimated as 22 per 100,000 farmers. The hand tools related injuries (8% of the total accidents) were non-fatal in nature. In spite of the enactment of legislation, the shortcomings in production and monitoring of the machinery in field use may be responsible for the high rate of accidents (e.g., 42 thresher accidents/1,000 mechanical threshers/year in southern India). Due to the lack of technical capability of the local artisans, adhering to safety and design standards is impractical to the implements fabricated in the rural areas. The analysis emphasizes that the effective safety and health management may be possible through legislative enabling of the local infra-structure, such as block development authority and primary health services, to permeate occupational health and safe work practices in the farming sector

  2. [An analysis of industrial accidents in the working field with a particular emphasis on repeated accidents].

    PubMed

    Wakisaka, I; Yanagihashi, T; Tomari, T; Sato, M

    1990-03-01

    The present study is based on an analysis of routinely submitted reports of occupational accidents experienced by the workers of industrial enterprises under the jurisdiction of Kagoshima Labor Standard Office during a 5-year period 1983 to 1987. Officially notified injuries serious enough to keep employees away from their job for work at least 4 days were utilized in this study. Data was classified so as to give an observed frequency distribution for workers having any specified number of accidents. Also, the accident rate which is an indicator of the risk of accident was compared among different occupations, between age groups and between the sexes. Results obtained are as follows; 1) For the combined total of 6,324 accident cases for 8 types of occupation (Construction, Transportation, Mining & Quarrying, Forestry, Food manufacture, Lumber & Woodcraft, Manufacturing industry and Other business), the number of those who had at least one accident was 6,098, of which 5,837 were injured only once, 208 twice, 21 three times and 2 four times. When occupation type was fixed, however, the number of workers having one, two, three and four times of accidents were 5,895, 182, 19 and 2, respectively. This suggests that some workers are likely to have experienced repeated accidents in more than one type of occupation.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:2131982

  3. Preliminary evaluation of the Accident Response Mobile Manipulation System for accident site salvage operations

    SciTech Connect

    Trujillo, J.M.; Morse, W.D.; Jones, D.P.

    1994-10-01

    This paper describes and evaluates operational experiences with the Accident Response Mobile Manipulation System (ARMMS) during simulated accident site salvage operations which might involve nuclear weapons. The ARMMS is based upon a teleoperated mobility platform with two Schilling Titan 7F Manipulators.

  4. The core paradox.

    NASA Technical Reports Server (NTRS)

    Kennedy, G. C.; Higgins, G. H.

    1973-01-01

    Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

  5. AN Core Analysis

    NASA Astrophysics Data System (ADS)

    Barbarino, Andrea; Tomatis, Daniele

    2014-06-01

    Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.

  6. Helicopter engine core noise

    NASA Astrophysics Data System (ADS)

    Vonglahn, U. H.

    1982-07-01

    Calculated engine core noise levels, based on NASA Lewis prediction procedures, for five representative helicopter engines are compared with measured total helicopter noise levels and ICAO helicopter noise certification requirements. Comparisons are made for level flyover and approach procedures. The measured noise levels are generally significantly greater than those predicted for the core noise levels, except for the Sikorsky S-61 and S-64 helicopters. However, the predicted engine core noise levels are generally at or within 3 dB of the ICAO noise rules. Consequently, helicopter engine core noise can be a significant contributor to the overall helicopter noise signature.

  7. Helicopter engine core noise

    NASA Technical Reports Server (NTRS)

    Vonglahn, U. H.

    1982-01-01

    Calculated engine core noise levels, based on NASA Lewis prediction procedures, for five representative helicopter engines are compared with measured total helicopter noise levels and ICAO helicopter noise certification requirements. Comparisons are made for level flyover and approach procedures. The measured noise levels are generally significantly greater than those predicted for the core noise levels, except for the Sikorsky S-61 and S-64 helicopters. However, the predicted engine core noise levels are generally at or within 3 dB of the ICAO noise rules. Consequently, helicopter engine core noise can be a significant contributor to the overall helicopter noise signature.

  8. Core Research Center

    USGS Publications Warehouse

    Hicks, Joshua; Adrian, Betty

    2009-01-01

    The Core Research Center (CRC) of the U.S. Geological Survey (USGS), located at the Denver Federal Center in Lakewood, Colo., currently houses rock core from more than 8,500 boreholes representing about 1.7 million feet of rock core from 35 States and cuttings from 54,000 boreholes representing 238 million feet of drilling in 28 States. Although most of the boreholes are located in the Rocky Mountain region, the geologic and geographic diversity of samples have helped the CRC become one of the largest and most heavily used public core repositories in the United States. Many of the boreholes represented in the collection were drilled for energy and mineral exploration, and many of the cores and cuttings were donated to the CRC by private companies in these industries. Some cores and cuttings were collected by the USGS along with other government agencies. Approximately one-half of the cores are slabbed and photographed. More than 18,000 thin sections and a large volume of analytical data from the cores and cuttings are also accessible. A growing collection of digital images of the cores are also becoming available on the CRC Web site Internet http://geology.cr.usgs.gov/crc/.

  9. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  10. Applicability of BWR SFD experiments and codes for advanced core component designs

    SciTech Connect

    Ott, L.J.

    1997-12-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories (SNL) in 1986, no experimental database existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper presents the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core on of SFD code.

  11. Single pilot IFR accident data analysis

    NASA Technical Reports Server (NTRS)

    Harris, D. F.; Morrisete, J. A.

    1982-01-01

    The aircraft accident data recorded and maintained by the National Transportation Safety Board for 1964 to 1979 were analyzed to determine what problems exist in the general aviation single pilot instrument flight rules environment. A previous study conducted in 1978 for the years 1964 to 1975 provided a basis for comparison. The purpose was to determine what changes, if any, have occurred in trends and cause-effect relationships reported in the earlier study. The increasing numbers have been tied to measures of activity to produce accident rates which in turn were analyzed in terms of change. Where anomalies or unusually high accident rates were encountered, further analysis was conducted to isolate pertinent patterns of cause factors and/or experience levels of involved pilots. The bulk of the effort addresses accidents in the landing phase of operations. A detailed analysis was performed on controlled/uncontrolled collisions and their unique attributes delineated. Estimates of day vs. night general aviation activity and accident rates were obtained.

  12. The three essentials for accident prevention.

    PubMed

    Eastman, Crystal

    2014-11-01

    This article was written by Crystal Eastman when she was Secretary of the New York Commission on Employers' Liability and Causes of Industrial Accidents, Unemployment, and Lack of Farm Labor. It was published in July of 1911, in Volume 38, Number 1 of the Annals of the American Academy of Political and Social Science, pages 98-107. The issue title was "Risks in Modern Industry." Eastman calls for the prevention of workplace accidents through three essentials: injury surveillance/reporting (with annual public reporting of the data); government enforcement of accident prevention laws, via departments with well-paid and well-trained officials and inspectors, fines that are high enough to be a deterrence to employers, and the power to have police shut down a factory if preventive measures are not installed; and a workers' compensation system-"a system of liability by which an employer can reduce his accident costs, not by hiring a more unscrupulous attorney and a more hard-hearted claim agent, but only by reducing his accidents." PMID:25261022

  13. NASA Medical Response to Human Spacecraft Accidents

    NASA Technical Reports Server (NTRS)

    Patlach, Robert

    2011-01-01

    This slide presentation reviews NASA's role in the response to spacecraft accidents that involve human fatalities or injuries. Particular attention is given to the work of the Mishap Investigation Team (MIT), the first response to the accidents and the interface to the accident investigation board. The MIT does not investigate the accident, but the objective of the MIT is to gather, guard, preserve and document the evidence. The primary medical objectives of the MIT is to receive, analyze, identify, and transport human remains, provide assistance in the recovery effort, and to provide family Casualty Coordinators with latest recovery information. The MIT while it does not determine the cause of the accident, it acts as the fact gathering arm of the Mishap Investigation Board (MIB), which when it is activated may chose to continue to use the MIT as its field investigation resource. The MIT membership and the specific responsibilities and tasks of the flight surgeon is reviewed. The current law establishing the process is also reviewed.

  14. Under-reporting of maritime accidents.

    PubMed

    Psarros, George; Skjong, Rolf; Eide, Magnus Strandmyr

    2010-03-01

    The majority of current maritime regulations has been developed following a reactive approach, often as ad-hoc response to serious accidents, and are characterised as being prescriptive leaving limited space for adapting equivalent solutions rather those described in the regulations. On the premise of providing a more proactive approach for the proposal or the evaluation of regulations, the Formal Safety Assessment (FSA) has been introduced. In the context of FSA, the analysis of accident data is considered to be very important for providing potential input on developing more balanced, proactive and cost-effective regulations. However, it has been argued that the validity of historical data may be undermined by uncertainties. This paper is aimed at showing evidence on serious under-reporting in accident databases, which can be considered as the main contributor to questioning the direct and uncritical use of historical data. By analysing the 10-year tanker accident data from the Lloyd's Register FairPlay (LRFP) and the Norwegian Maritime Directorate (NMD) for vessels registered in Norway, it is found that the reporting performance has an upper bound of 41% for NMD and 30% for LRFP. Furthermore, based on comparison between LRFP data and self-assessment by Flag States, it is seen that accidents reported by the Flag States are also incomplete. PMID:20159087

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

    SciTech Connect

    1998-03-01

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

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

    SciTech Connect

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. )

    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.

  17. SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS

    SciTech Connect

    Brad J. Merrill; Shannon M Bragg-Sitton

    2013-09-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) Light Water Reactor (LWR) Sustainability Program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. The Fuels Pathway within this program focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement (e.g. fully ceramic cladding). The DOE-NE Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) is also conducting research on materials for advanced, accident tolerant fuels and cladding for application in operating LWRs. To aide in this assessment, a silicon carbide (SiC) version of the MELCOR code was developed by substituting SiC in place of Zircaloy in MELCOR’s reactor core oxidation and material property routines. The purpose of this development effort is to provide a numerical capability for estimating the safety advantages of replacing Zr-alloy components in LWRs with SiC components. This modified version of the MELCOR code was applied to the Three Mile Island (TMI-2) plant accident. While the results are considered preliminary, SiC cladding showed a dramatic safety advantage over Zircaloy cladding during this accident.

  18. Reconceptualising Core Skills

    ERIC Educational Resources Information Center

    Canning, Roy

    2007-01-01

    The paper provides an analysis of Core Skill policy and practice in the UK. The author presents a conceptual basis for re-thinking generic Core Skills within educational approaches in teaching and learning. The discussion looks at whether universal notions of generic skills are appropriate when considering post-compulsory pedagogic approaches to…

  19. The Common Core.

    ERIC Educational Resources Information Center

    Boyer, Ernest L.

    Current curricula in institutions of higher education are criticized in this speech for their lack of a common core of education. Several possibilities for developing such a common core include education centered around our common heritage and the challenges of the present. It is suggested that all students must be introduced to the events,…

  20. Making an Ice Core.

    ERIC Educational Resources Information Center

    Kopaska-Merkel, David C.

    1995-01-01

    Explains an activity in which students construct a simulated ice core. Materials required include only a freezer, food coloring, a bottle, and water. This hands-on exercise demonstrates how a glacier is formed, how ice cores are studied, and the nature of precision and accuracy in measurement. Suitable for grades three through eight. (Author/PVD)

  1. Iowa Core Annual Report

    ERIC Educational Resources Information Center

    Iowa Department of Education, 2015

    2015-01-01

    One central component of a great school system is a clear set of expectations, or standards, that educators help all students reach. In Iowa, that effort is known as the Iowa Core. The Iowa Core represents the statewide academic standards, which describe what students should know and be able to do in math, science, English language arts, and…

  2. CORE - Performance Feedback System

    Energy Science and Technology Software Center (ESTSC)

    2009-10-02

    CORE is an architecture to bridge the gaps between disparate data integration and delivery of disparate information visualization. The CORE Technology Program includes a suite of tools and user-centered staff that can facilitate rapid delivery of a deployable integrated information to users.

  3. Core Skills in Action.

    ERIC Educational Resources Information Center

    Further Education Unit, London (England).

    This bulletin provides an update on current developments in core skills in further education. Section 1 contains information about the Further Education Unit's (FEU's) Core Skills Post-16 project, in which colleges are testing principles that underpin all good quality learning programs. Important findings and examples are outlined under the five…

  4. NFE Core Bibliographies.

    ERIC Educational Resources Information Center

    Michigan State Univ., East Lansing. Inst. for International Studies in Education.

    This collection of core bibliographies, which expands on an initial bibliography published in 1979 of the core resources housed in the Non-Formal Education Information Center at Michigan State University, comprises a basic stock of materials on nonformal education and women in development that have been contributed by development planners,…

  5. Ice Core Investigations

    ERIC Educational Resources Information Center

    Krim, Jessica; Brody, Michael

    2008-01-01

    What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…

  6. CFD Analyses of Air-Ingress Accident for VHTRs

    NASA Astrophysics Data System (ADS)

    Ham, Tae Kyu

    The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air

  7. CFD Analyses of Air-Ingress Accident for VHTRs

    NASA Astrophysics Data System (ADS)

    Ham, Tae Kyu

    The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air

  8. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect

    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.

  9. A Review of Criticality Accidents 2000 Revision

    SciTech Connect

    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.

  10. Predicting road accidents: Structural time series approach

    NASA Astrophysics Data System (ADS)

    Junus, Noor Wahida Md; Ismail, Mohd Tahir

    2014-07-01

    In this paper, the model for occurrence of road accidents in Malaysia between the years of 1970 to 2010 was developed and throughout this model the number of road accidents have been predicted by using the structural time series approach. The models are developed by using stepwise method and the residual of each step has been analyzed. The accuracy of the model is analyzed by using the mean absolute percentage error (MAPE) and the best model is chosen based on the smallest Akaike information criterion (AIC) value. A structural time series approach found that local linear trend model is the best model to represent the road accidents. This model allows level and slope component to be varied over time. In addition, this approach also provides useful information on improving the conventional time series method.

  11. Allometric scaling and accidents at work

    PubMed Central

    Cempel, Czesław; Tabaszewski, Maciej; Ordysiński, Szymon

    2016-01-01

    Allometry is the knowledge concerning relations between the features of some beings, like animals, or cities. For example, the daily energy rate is proportional to a mass of mammals rise of 3/4. This way of thinking has spread quickly from biology to many areas of research concerned with sociotechnical systems. It was revealed that the number of innovations, patents or heavy crimes rises as social interaction increases in a bigger city, while other urban indexes such as suicides decrease with social interaction. Enterprise is also a sociotechnical system, where social interaction and accidents at work take place. Therefore, do these interactions increase the number of accidents at work or, on the contrary, are they reduction-driving components? This article tries to catch such links and assess the allometric exponent between the number of accidents at work and the number of employees in an enterprise. PMID:26655044

  12. Reconfigurable mobile manipulation for accident response

    SciTech Connect

    ANDERSON,ROBERT J.; MORSE,WILLIAM D.; SHIREY,DAVID L.; CDEBACA,DANIEL M.; HOFFMAN JR.,JOHN P.; LUCY,WILLIAM E.

    2000-06-06

    The need for a telerobotic vehicle with hazard sensing and integral manipulation capabilities has been identified for use in transportation accidents where nuclear weapons are involved. The Accident Response Mobile Manipulation System (ARMMS) platform has been developed to provide remote dexterous manipulation and hazard sensing for the Accident Response Group (ARG) at Sandia National Laboratories. The ARMMS' mobility platform is a military HMMWV [High Mobility Multipurpose Wheeled Vehicle] that is teleoperated over RF or Fiber Optic communication channels. ARMMS is equipped with two high strength Schilling Titan II manipulators and a suite of hazardous gas and radiation sensors. Recently, a modular telerobotic control architecture call SMART (Sandia Modular Architecture for Robotic and Teleoperation) has been applied to ARMMS. SMART enables input devices and many system behaviors to be rapidly configured in the field for specific mission needs. This paper summarizes current SMART developments applied to ARMMS.

  13. Agricultural implications of the Fukushima nuclear accident

    PubMed Central

    Nakanishi, Tomoko M.

    2016-01-01

    More than 4 years has passed since the accident at the Fukushima Nuclear Power Plant. Immediately after the accident, 40 to 50 academic staff of the Graduate School of Agricultural and Life Sciences at the University of Tokyo created an independent team to monitor the behavior of the radioactive materials in the field and their effects on agricultural farm lands, forests, rivers, animals, etc. When the radioactive nuclides from the nuclear power plant fell, they were instantly adsorbed at the site where they first touched; consequently, the fallout was found as scattered spots on the surface of anything that was exposed to the air at the time of the accident. The adsorption has become stronger over time, so the radioactive nuclides are now difficult to remove. The findings of our study regarding the wide range of effects on agricultural fields are summarized in this report. PMID:27538845

  14. Agricultural implications of the Fukushima nuclear accident.

    PubMed

    Nakanishi, Tomoko M

    2016-08-01

    More than 4 years has passed since the accident at the Fukushima Nuclear Power Plant. Immediately after the accident, 40 to 50 academic staff of the Graduate School of Agricultural and Life Sciences at the University of Tokyo created an independent team to monitor the behavior of the radioactive materials in the field and their effects on agricultural farm lands, forests, rivers, animals, etc. When the radioactive nuclides from the nuclear power plant fell, they were instantly adsorbed at the site where they first touched; consequently, the fallout was found as scattered spots on the surface of anything that was exposed to the air at the time of the accident. The adsorption has become stronger over time, so the radioactive nuclides are now difficult to remove. The findings of our study regarding the wide range of effects on agricultural fields are summarized in this report. PMID:27538845

  15. Allometric scaling and accidents at work.

    PubMed

    Cempel, Czesław; Tabaszewski, Maciej; Ordysiński, Szymon

    2016-01-01

    Allometry is the knowledge concerning relations between the features of some beings, like animals, or cities. For example, the daily energy rate is proportional to a mass of mammals rise of 3/4. This way of thinking has spread quickly from biology to many areas of research concerned with sociotechnical systems. It was revealed that the number of innovations, patents or heavy crimes rises as social interaction increases in a bigger city, while other urban indexes such as suicides decrease with social interaction. Enterprise is also a sociotechnical system, where social interaction and accidents at work take place. Therefore, do these interactions increase the number of accidents at work or, on the contrary, are they reduction-driving components? This article tries to catch such links and assess the allometric exponent between the number of accidents at work and the number of employees in an enterprise. PMID:26655044

  16. Personality characteristics of the child accident repeater

    PubMed Central

    Manheimer, Dean I; Mellinger, Glen D

    1997-01-01

    From 8874 boys and girls aged 4 to 18 in Berkeley-Oakland, California, 684 were selected to represent high-, intermediateor low-accident-liability children, based on records of medically attended injuries. Using data from intensive interviews with mothers supplemented with school records, we found a statistically significant relation between accident liability and indexes of extraversion, daring, roughhousing, and other traits tending to expose children to hazards. Similar relations held for traits such as poor discipline, aggressiveness toward peers, and, for girls, attention-seeking, which compete with the child's ability to cope with hazards. Other traits that may impair ability to cope with hazards were also found to be related to accident liability (impulsivity, carelessness, and unreliability) as were several indexes denoting maladjustment. PMID:9213161

  17. Structural aspects of the Chernobyl accident

    SciTech Connect

    Murray, R.C.; Cummings, G.E.

    1988-09-02

    On April 26, 1986 the world's worst nuclear power plant accident occurred at the Unit 4 of the Chernobyl Nuclear Power Station in the USSR. This paper presents a discussion of the design of the Chernobyl Power Plant, the sequence of events that led to the accident and the damage caused by the resulting explosion. The structural design features that contributed to the accident and resulting damage will be highlighted. Photographs and sketches obtained from various worldwide news agencies will be shown to try and gain a perspective of the extent of the damage. The aftermath, clean-up, and current situation will be discussed and the important lessons learned for the structural engineer will be presented. 15 refs., 10 figs.

  18. Lunar Core and Tides

    NASA Technical Reports Server (NTRS)

    Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

    2004-01-01

    Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2,3,4] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening, which in the past has been marginal but improving [3,4,5], now seems significant. Direct detection of the core moment has not yet been achieved.

  19. Internal core tightener

    DOEpatents

    Brynsvold, Glen V.; Snyder, Jr., Harold J.

    1976-06-22

    An internal core tightener which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a "fixed" outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change.

  20. Use of artificial intelligence in severe accident diagnosis for PWRs

    SciTech Connect

    Wu, Zheng; Okrent, D.; Kastenberg, W.E.

    1995-12-31

    A combination approach of an expert system and neural networks is used to implement a prototype severe accident diagnostic system which would monitor the progression of the severe accident and provide necessary plant status information to assist the plant staff in accident management during the accident. The station blackout accident in a pressurized water reactor (PWR) is used as the study case. The current phase of research focus is on distinguishing different primary system failure modes and following the accident transient before and up to vessel breach.

  1. A study on industrial accident rate forecasting and program development of estimated zero accident time in Korea.

    PubMed

    Kim, Tae-gu; Kang, Young-sig; Lee, Hyung-won

    2011-01-01

    To begin a zero accident campaign for industry, the first thing is to estimate the industrial accident rate and the zero accident time systematically. This paper considers the social and technical change of the business environment after beginning the zero accident campaign through quantitative time series analysis methods. These methods include sum of squared errors (SSE), regression analysis method (RAM), exponential smoothing method (ESM), double exponential smoothing method (DESM), auto-regressive integrated moving average (ARIMA) model, and the proposed analytic function method (AFM). The program is developed to estimate the accident rate, zero accident time and achievement probability of an efficient industrial environment. In this paper, MFC (Microsoft Foundation Class) software of Visual Studio 2008 was used to develop a zero accident program. The results of this paper will provide major information for industrial accident prevention and be an important part of stimulating the zero accident campaign within all industrial environments. PMID:20823633

  2. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-09-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios.

  3. Potassium iodide for thyroid blockade in a reactor accident: administrative policies that govern its use.

    PubMed

    Becker, D V; Zanzonico, P

    1997-04-01

    structures with large steel and concrete shells and multiple redundancy of core cooling mechanisms. These successfully prevented the release of major amounts of radionuclides in the Three Mile Island partial loss-of-primary coolant accident in 1979. The Chernobyl accident, in a different type of reactor that is common in Eastern Europe, did not have effective outer shell containment and released almost 50 MCi of 131I compared to the 20 Ci of 131I released at Three Mile Island. Such accidents have precipitated extensive re-evaluation of the design and safety devices of all operating reactors. However, a major contributing factor to the accidents was human error and considerable efforts must be made to train plant operators so they have a better understanding of reactor operation and use of safety mechanisms. PMID:9133683

  4. Variant 22: Spatially-Dependent: Transient Processes in MOX Fueled Core

    SciTech Connect

    Pavlovichev, A.M.

    2001-09-28

    This work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, and to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: (1) Central control rod ejection by pressure drop caused by destroying of the moving mechanism cover. (2) Overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve. (3) The boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop. These accidents have been applied to: (1) Uranium reference core that is the so-called Advanced VVER-1000 core with Zirconium fuel pins claddings and guide tubes. A number of assemblies contained 18 boron BPRs while first year operating. (2) MOX core with about 30% MOX fuel. At a solving it was supposed that MOX-fuel thermophysical characteristics are identical to uranium fuel ones. The calculations were carried out with the help of the program NOSTRA/1/, simulating VVER dynamics that is briefly described in Chapter 1. Chapter 3 contains the description of reference Uranium and MOX cores that are used in calculations. The neutronics calculations of MOX core with about 30% MOX fuel are named ''Variant 2 1''. Chapters 4-6 contain the calculational results of three above mentioned benchmark accidents that compose in a whole the ''Variant 22''.

  5. Previous convictions or accidents and the risk of subsequent accidents of older drivers.

    PubMed

    Daigneault, Geneviève; Joly, Pierre; Frigon, Jean-Yves

    2002-03-01

    The over-involvement of elderly drivers in collisions has a potentially adverse effect on highway safety. The question for most experts in traffic research is whether we can predict the individual risk of accidents and which variables are the best predictors, especially for this population. For a better understanding of the elderly drivers' problems, this study aimed to describe the most common types of accidents in the elderly population of drivers living in Quebec (> or = 65 years of age). The second objective of the study was to analyse the relationship between previous accidents or convictions and the risk of subsequent accidents. The results show that: (1) elderly drivers are characterised by error accidents involving more than one car, especially at intersections, (2) prior accidents are a better predictor for accident risk than prior convictions and (3) these trends steadily increase with each age group (drivers 65 years old to 80 years or more). The results are discussed in relation to the literature on risk behaviour of the elderly drivers. PMID:11829296

  6. French policy for managing the post-accident phase of a nuclear accident.

    PubMed

    Gallay, F; Godet, J L; Niel, J C

    2015-06-01

    In 2005, at the request of the French Government, the Nuclear Safety Authority (ASN) established a Steering Committee for the Management of the Post-Accident Phase of a Nuclear Accident or a Radiological Emergency, with the objective of establishing a policy framework. Under the supervision of ASN, this Committee, involving several tens of experts from different backgrounds (e.g. relevant ministerial offices, expert agencies, local information commissions around nuclear installations, non-governmental organisations, elected officials, licensees, and international experts), developed a number of recommendations over a 7-year period. First published in November 2012, these recommendations cover the immediate post-emergency situation, and the transition and longer-term periods of the post-accident phase in the case of medium-scale nuclear accidents causing short-term radioactive release (less than 24 h) that might occur at French nuclear facilities. They also apply to actions to be undertaken in the event of accidents during the transportation of radioactive materials. These recommendations are an important first step in preparation for the management of a post-accident situation in France in the case of a nuclear accident. PMID:25915552

  7. Sodium hypochlorite accident in a pediatric patient.

    PubMed

    Klein, Ulrich; Kleier, Donald J

    2013-01-01

    A sodium hypochlorite accident is a rare event in adults, but even more so in children. The purposes of this paper were to: report the case of a one-year, 10-month-old toddler who was treated under general anesthesia for early childhood caries, incurred a sodium hypochlorite accident following attempted pulpectomy in his primary maxillary central incisors, but made a full recovery without any sequelae after a typical four to six weeks course of disease; review the pertinent literature; and give recommendations on how to minimize the potential occurrence of such incidents. PMID:24553278

  8. Dose estimates from the Chernobyl accident

    SciTech Connect

    Lange, R.; Dickerson, M.H.; Gudiksen, P.H.

    1987-11-01

    The Lawrence Livermore National Laboratory Atmospheric Release Advisory Capability (ARAC) responded to the Chernobyl nuclear reactor accident in the Soviet Union by utilizing long-range atmospheric dispersion modeling to estimate the amount of radioactivity released (source term) and the radiation dose distribution due to exposure to the radioactive cloud over Europe and the Northern Hemisphere. In later assessments, after the release of data on the accident by the Soviet Union, the ARAC team used their mesoscale to regional scale model to focus in on the radiation dose distribution within the Soviet Union and the vicinity of the Chernobyl plant. 22 refs., 5 figs., 5 tabs.

  9. Lessons learned from early criticality accidents

    SciTech Connect

    Malenfant, R.E.

    1996-06-01

    Four accidents involving the approach to criticality occurred during the period July, 1945, through May, 1996. These have been described in the format of the OPERATING EXPERIENCE WEEKLY SUMMARY which is distributed by the Office of Nuclear and Facility Safety. Although the lessons learned have been incorporated in standards, codes, and formal procedures during the last fifty years, this is their first presentation in this format. It is particularly appropriate that they be presented in the forum of the Nuclear Criticality Technology Safety Project Workshop closest to the fiftieth anniversary of the last of the four accidents, and that which was most instrumental in demonstrating the need to incorporate lessons learned.

  10. The medical investigation of airship accidents.

    PubMed

    Stahl, C J; McMeekin, R R; Ruehle, C J; Canik, J J

    1988-07-01

    A review of the autopsy reports for 18 of 21 victims in 3 of the 4 nonrigid Navy airship accidents during the period 1955 to 1966 revealed that the patterns of injury, complicated by postcrash entrapment, immersion, or fire, are similar to the injuries observed in the low-speed, low-altitude crashes of rigid airships and of light aircraft. With the renewed interest in the development of airships for military purposes, there is a need for improved design related to crashworthiness and to aircrew habitability, safety, restraint, and egress in order to enhance the chance for survival in the event of an accident. PMID:3171506

  11. Accident analysis for US fast burst reactors

    SciTech Connect

    Paternoster, R.; Flanders, M.; Kazi, H.

    1994-09-01

    In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards.

  12. Czech Republic 20 years after Chernobyl accident.

    PubMed

    Rosina, Jozef; Kvasnák, Eugen; Suta, Daniel; Kostrhun, Tomás; Drábová, Dana

    2008-01-01

    The territory of the Czech Republic was contaminated as a result of the breakdown in the Chernobyl nuclear power plant in 1986. The Czech population received low doses of ionising radiation which, though it could not cause a deterministic impact, could have had stochastic effects expressed in the years following the accident. Twenty years after the accident is a long enough time to assess its stochastic effects, primarily tumours and genetic impairment. The moderate amount of radioactive fallout received by the Czech population in 1986 increased thyroid cancer in the following years; on the other hand, no obvious genetic impact was found. PMID:18375464

  13. Columbia Accident Investigation Board Report. Volume Four

    NASA Technical Reports Server (NTRS)

    Gehmann, H. W.; Barry, J. L.; Deal, D. W.; Hallock, J. N.; Hess, K. W.

    2003-01-01

    This is Volume Four of a set of six reports produced by NASA and other organizations which were provided to the Columbia Accident Investigation Board (CAIB) in support of its inquiry into the February 1, 2003 destruction of the Space Shuttle Columbia. The Technical Documents included in this volume are: Appendix F.1 Water Absorption by Foam; Appendix F.2 Follow the TPS; Appendix F.3 MADS Sensor Data; Appendix F.4 ET Cryoinsulation; Appendix F.5 Space Shuttle STS-107 Columbia Accident Investigation, and External Tank Working Group Final Report - Volume 1.

  14. LESSONS LEARNED FROM A RECENT LASER ACCIDENT

    SciTech Connect

    Woods, Michael; /SLAC

    2011-01-26

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

  15. Procedural accidents: an online study guide.

    PubMed

    2008-05-01

    The Editorial Board of the Journal of Endodontics has developed a literature-based study guide of topical areas related to endodontics. This study guide is intended to give the reader a focused review of the essential endodontic literature and does not cite all possible articles related to each topic. Although citing all articles would be comprehensive, it would defeat the idea of a study guide. This section will cover procedural accidents, including perforation repairs, separated instrument, and accidents with irrigants and endodontic materials. PMID:18457710

  16. 34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES CORES THAT ARE NOT MADE ON HEATED OR COLD BOX CORE MACHINES, TO SET BINDING AGENTS MIXED WITH THE SAND CREATING CORES HARD ENOUGH TO WITHSTAND THE FLOW OF MOLTEN IRON INSIDE A MOLD. - Stockham Pipe & Fittings Company, Grey Iron Foundry, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  17. World commercial aircraft accidents. Second edition, 1946--1992

    SciTech Connect

    Kimura, C.Y.

    1993-01-01

    This report is a compilation of all accidents world-wide involving aircraft in commercial service which resulted in the loss of the airframe or one or more fatality, or both. This information has been gathered in order to present a complete inventory of commercial aircraft accidents. Events involving military action, sabotage, terrorist bombings, hijackings, suicides, and industrial ground accidents are included within this list. Included are: accidents involving world commercial jet aircraft, world commercial turboprop aircraft, world commercial pistonprop aircraft with four or more engines and world commercial pistonprop aircraft with two or three engines from 1946 to 1992. Each accident is presented with information in the following categories: date of the accident, airline and its flight numbers, type of flight, type of aircraft, aircraft registration number, construction number/manufacturers serial number, aircraft damage, accident flight phase, accident location, number of fatalities, number of occupants, cause, remarks, or description (brief) of the accident, and finally references used. The sixth chapter presents a summary of the world commercial aircraft accidents by major aircraft class (e.g. jet, turboprop, and pistonprop) and by flight phase. The seventh chapter presents several special studies including a list of world commercial aircraft accidents for all aircraft types with 100 or more fatalities in order of decreasing number of fatalities, a list of collision accidents involving commercial aircrafts, and a list of world commercial aircraft accidents for all aircraft types involving military action, sabotage, terrorist bombings, and hijackings.

  18. Three Mile Island Unit-2 core status summary: a basis for tool development for reactor disassembly and defueling

    SciTech Connect

    Croucher, D.W.

    1981-05-01

    The accident at Three Mile Island Unit-2 (TMI-2) on March 28, 1979 caused extensive damage to the core. A variety of analyses were performed using three general approaches to determine the extent of core damage. First, thermal-hydraulic events were reconstructed using available data, thermal-hydraulic principles, and computer analyses. Second, determinations of the hydrogen generated yielded estimates of the amount of zircaloy oxidized and embrittled. Third, the type and quantity of fission products released during the accident were used to estimate the location of core damage and the fuel temperatures which were achieved. Uncertainties exist in each type of determination due to the equivocal nature of the data. This paper reviews and summarizes the core damage assessments which have been made, identifies the minimum and maximum bounds of damage, and establishes a reference description for the current status of the core.

  19. Multiple Core Galaxies

    NASA Technical Reports Server (NTRS)

    Miller, R.H.; Morrison, David (Technical Monitor)

    1994-01-01

    Nuclei of galaxies often show complicated density structures and perplexing kinematic signatures. In the past we have reported numerical experiments indicating a natural tendency for galaxies to show nuclei offset with respect to nearby isophotes and for the nucleus to have a radial velocity different from the galaxy's systemic velocity. Other experiments show normal mode oscillations in galaxies with large amplitudes. These oscillations do not damp appreciably over a Hubble time. The common thread running through all these is that galaxies often show evidence of ringing, bouncing, or sloshing around in unexpected ways, even though they have not been disturbed by any external event. Recent observational evidence shows yet another phenomenon indicating the dynamical complexity of central regions of galaxies: multiple cores (M31, Markarian 315 and 463 for example). These systems can hardly be static. We noted long-lived multiple core systems in galaxies in numerical experiments some years ago, and we have more recently followed up with a series of experiments on multiple core galaxies, starting with two cores. The relevant parameters are the energy in the orbiting clumps, their relative.masses, the (local) strength of the potential well representing the parent galaxy, and the number of cores. We have studied the dependence of the merger rates and the nature of the final merger product on these parameters. Individual cores survive much longer in stronger background potentials. Cores can survive for a substantial fraction of a Hubble time if they travel on reasonable orbits.

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

    SciTech Connect

    Gonnet, M.

    2012-07-01

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

  1. Loss of DHR sequences at Browns Ferry Unit One - accident-sequence analysis

    SciTech Connect

    Cook, D.H.; Grene, S.R.; Harrington, R.M.; Hodge, S.A.

    1983-05-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated loss of decay heat removal (DHR) capability following scram from full power with the power conversion system unavailable. In accident sequences without DHR capability, the residual heat removal (RHR) system functions of pressure suppression pool cooling and reactor vessel shutdown cooling are unavailable. Consequently, all decay heat energy is stored in the pressure suppression pool with a concomitant increase in pool temperature and primary containment pressure. With the assumption that DHR capability is not regained during the lengthy course of this accident sequence, the containment ultimately fails by overpressurization. Although unlikely, this catastrophic failure might lead to loss of the ability to inject cooling water into the reactor vessel, causing subsequent core uncovery and meltdown. The timing of these events and the effective mitigating actions that might be taken by the operator are discussed in this report.

  2. Advanced accident sequence precursor analysis level 1 models

    SciTech Connect

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O.

    1996-03-01

    INEL has been involved in the development of plant-specific Accident Sequence Precursor (ASP) models for the past two years. These models were developed for use with the SAPHIRE suite of PRA computer codes. They contained event tree/linked fault tree Level 1 risk models for the following initiating events: general transient, loss-of-offsite-power, steam generator tube rupture, small loss-of-coolant-accident, and anticipated transient without scram. Early in 1995 the ASP models were revised based on review comments from the NRC and an independent peer review. These models were released as Revision 1. The Office of Nuclear Regulatory Research has sponsored several projects at the INEL this fiscal year to further enhance the capabilities of the ASP models. Revision 2 models incorporates more detailed plant information into the models concerning plant response to station blackout conditions, information on battery life, and other unique features gleaned from an Office of Nuclear Reactor Regulation quick review of the Individual Plant Examination submittals. These models are currently being delivered to the NRC as they are completed. A related project is a feasibility study and model development of low power/shutdown (LP/SD) and external event extensions to the ASP models. This project will establish criteria for selection of LP/SD and external initiator operational events for analysis within the ASP program. Prototype models for each pertinent initiating event (loss of shutdown cooling, loss of inventory control, fire, flood, seismic, etc.) will be developed. A third project concerns development of enhancements to SAPHIRE. In relation to the ASP program, a new SAPHIRE module, GEM, was developed as a specific user interface for performing ASP evaluations. This module greatly simplifies the analysis process for determining the conditional core damage probability for a given combination of initiating events and equipment failures or degradations.

  3. Boson core compressibility

    NASA Astrophysics Data System (ADS)

    Khorramzadeh, Y.; Lin, Fei; Scarola, V. W.

    2012-04-01

    Strongly interacting atoms trapped in optical lattices can be used to explore phase diagrams of Hubbard models. Spatial inhomogeneity due to trapping typically obscures distinguishing observables. We propose that measures using boson double occupancy avoid trapping effects to reveal two key correlation functions. We define a boson core compressibility and core superfluid stiffness in terms of double occupancy. We use quantum Monte Carlo on the Bose-Hubbard model to empirically show that these quantities intrinsically eliminate edge effects to reveal correlations near the trap center. The boson core compressibility offers a generally applicable tool that can be used to experimentally map out phase transitions between compressible and incompressible states.

  4. Global Core Plasma Model

    NASA Technical Reports Server (NTRS)

    Gallagher, Dennis L.; Craven, P. D.; Comfort, R. H.

    1999-01-01

    Abstract. The Global Core Plasma Model (GCPM) provides, empirically derived, core plasma density as a function of geomagnetic and solar conditions throughout the inner magnetosphere. It is continuous in value and gradient and is composed of separate models for the ionosphere, the plasmasphere, the plasmapause, the trough, and the polar cap. The relative composition of plasmaspheric H+, He+, and O+ is included in the GCPM. A blunt plasmaspheric bulge and rotation of the bulge with changing geomagnetic conditions is included. The GCPM is an amalgam of density models, intended to serve as a framework for continued improvement as new measurements become available and are used to characterize core plasma density, composition, and temperature.

  5. Core shroud corner joints

    DOEpatents

    Gilmore, Charles B.; Forsyth, David R.

    2013-09-10

    A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

  6. Remote control continuous mining machine crushing accident data study

    SciTech Connect

    2006-05-11

    A committee was formed to identify norms and trends in remote control continuous miner crushing accidents as part of US MSHA's efforts to reduce and eliminate these types of accidents. The committee was tasked with collecting, reviewing, and evaluating remote control accident data to identify significant factors that could possibly contribute to remote control accidents. The report identifies that these types of accidents commonly happen to experienced miners during routine mining activities, with the majority occurring while moving the miner from one face to another (place changing). Another common aspect of the accidents is that many of the victims are newly employed at the mine where the accident occurred. Training all employees to stay outside the turning radius of an energized remote control continuous miner, establishing this as a safe operating procedure, and consistently enforcing this practice among miners will reduce these types of accidents. 10 figs., 5 tabs., 7 apps.

  7. Investigation of shipping accident injury severity and mortality.

    PubMed

    Weng, Jinxian; Yang, Dong

    2015-03-01

    Shipping movements are operated in a complex and high-risk environment. Fatal shipping accidents are the nightmares of seafarers. With ten years' worldwide ship accident data, this study develops a binary logistic regression model and a zero-truncated binomial regression model to predict the probability of fatal shipping accidents and corresponding mortalities. The model results show that both the probability of fatal accidents and mortalities are greater for collision, fire/explosion, contact, grounding, sinking accidents occurred in adverse weather conditions and darkness conditions. Sinking has the largest effects on the increment of fatal accident probability and mortalities. The results also show that the bigger number of mortalities is associated with shipping accidents occurred far away from the coastal area/harbor/port. In addition, cruise ships are found to have more mortalities than non-cruise ships. The results of this study are beneficial for policy-makers in proposing efficient strategies to prevent fatal shipping accidents. PMID:25617776

  8. Cross-analysis of hazmat road accidents using multiple databases.

    PubMed

    Trépanier, Martin; Leroux, Marie-Hélène; de Marcellis-Warin, Nathalie

    2009-11-01

    Road selection for hazardous materials transportation relies heavily on risk analysis. With risk being generally expressed as a product of the probability of occurrence and the expected consequence, one will understand that risk analysis is data intensive. However, various authors have noticed the lack of statistical reliability of hazmat accident databases due to the systematic underreporting of such events. Also, official accident databases alone are not always providing all the information required (economical impact, road conditions, etc.). In this paper, we attempt to integrate many data sources to analyze hazmat accidents in the province of Quebec, Canada. Databases on dangerous goods accidents, road accidents and work accidents were cross-analyzed. Results show that accidents can hardly be matched and that these databases suffer from underreporting. Police records seem to have better coverage than official records maintained by hazmat authorities. Serious accidents are missing from government's official databases (some involving deaths or major spills) even though their declaration is mandatory. PMID:19819367

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

    SciTech Connect

    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. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a... information: (1) Date, time, and approximate duration of the release; (2) Chemical(s) released; (3)...

  11. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a... information: (1) Date, time, and approximate duration of the release; (2) Chemical(s) released; (3)...

  12. A general approach to critical infrastructure accident consequences analysis

    NASA Astrophysics Data System (ADS)

    Bogalecka, Magda; Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

    2016-06-01

    The probabilistic general model of critical infrastructure accident consequences including the process of the models of initiating events generated by its accident, the process of environment threats and the process of environment degradation is presented.

  13. A description of the general aviation fixed wing accident

    NASA Technical Reports Server (NTRS)

    Hall, D. S.

    1981-01-01

    The Emergency Local Transmitter (ELT) is a radio transmitter with a self-contained power source designed to provide notification of and homing to aircraft accident sites. The Crash Research Institute has monitored general aviation fixed-wing accidents in the United States and in Canada and has found that: (1) the ELT was destroyed in approximately 25% of all fatal accidents; (2) the ELT activated in about 62% of the fatal accidents, 69% of the fatal with survivors accidents, almost 80% of the serious accidents and about 57% of the minor accidents; (3) in fatal accidents the aircraft sections least likely to be destroyed are the vertical and horizontal tail surfaces; (4) antenna cable disconnection and antenna breakage caused failure to transmit usable signals; and (5) initial alerting control occurred in nearly half of the situations where the ELT aided in search.

  14. 36 CFR 9.46 - Accidents and fires.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... outlined in 36 CFR 2.17, but does not relieve persons from the responsibility of making any other accident... MINERALS MANAGEMENT Non-Federal Oil and Gas Rights § 9.46 Accidents and fires. The operator shall...

  15. 36 CFR 9.46 - Accidents and fires.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... outlined in 36 CFR 2.17, but does not relieve persons from the responsibility of making any other accident... MINERALS MANAGEMENT Non-Federal Oil and Gas Rights § 9.46 Accidents and fires. The operator shall...

  16. School Bus Accidents: Reducing Incidents and Injuries

    ERIC Educational Resources Information Center

    Mahoney, Daniel

    2009-01-01

    The number of children injured in nonfatal school bus accidents annually is more than double the number previously estimated. In Ohio alone, approximately 20,800 children younger than 18 were occupants of school buses that were involved in crashes in 2003 and 2004 (McGeehan 2007). Among those children, most had minor or no injuries. However, there…

  17. Speed Variance and Its Influence on Accidents.

    ERIC Educational Resources Information Center

    Garber, Nicholas J.; Gadirau, Ravi

    A study was conducted to investigate the traffic engineering factors that influence speed variance and to determine to what extent speed variance affects accident rates. Detailed analyses were carried out to relate speed variance with posted speed limit, design speeds, and other traffic variables. The major factor identified was the difference…

  18. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ..., DEPARTMENT OF TRANSPORTATION RAIL FIXED GUIDEWAY SYSTEMS; STATE SAFETY OVERSIGHT Role of the State Oversight Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle...

  19. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ..., DEPARTMENT OF TRANSPORTATION RAIL FIXED GUIDEWAY SYSTEMS; STATE SAFETY OVERSIGHT Role of the State Oversight Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle...

  20. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ..., DEPARTMENT OF TRANSPORTATION RAIL FIXED GUIDEWAY SYSTEMS; STATE SAFETY OVERSIGHT Role of the State Oversight Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle...

  1. Tragic Car Accident Involves ESO Employees

    NASA Astrophysics Data System (ADS)

    2000-06-01

    Saturday, May 27, turned into a tragic day for ESO. The team installing TIMMI2 at La Silla, went on an excursion to the Elqui valley, 70 km east of the city of La Serena and suffered a serious car accident, crashing against another car driving from the opposite direction.

  2. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 4 2010-07-01 2010-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  3. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 4 2011-07-01 2011-07-01 false Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  4. [Risk of accidents for a Parisian pedestrian].

    PubMed

    Vayre, P

    2001-12-01

    The risk of accident for a parisian pedestrian is of reduced frequency and of moderate severity (terrorism and natural disasters not included) according to the statistics of 1999. This is due to security measures and the excellent organisation of assistance to wounded persons. The improvement in protection of pedestrians, even the disabled and elderly persons will soon make Paris the best protected megalopolis of our continent. Pedestrians are involved in 28% of accidents on public streets with a mortality of 1.2%, with serious injuries in 12.6% of the cases. Injuries without collision occur in 57% of the cases. 18% of the wounded older than 75 years with 1.8% of deaths and 66% of severe injury. Accidents related to public transportation represent a very slight risk of 0.00044%. By modification of the concept of transportation and its materials over a five year period there is a decrease in the number of the victims (-3.75%), in spite of the increase of accidents (3.5%). For lone pedestrian being victims of aggression in public places, after a decrease of 10% in 1999, there is in 2001 an increase which raises the question of police proximity again. Over a five year period the RATP (subway authority) has noticed a 14.77% decrease of thefts and a 54.78% decrease in attacks... This makes the Parisian railnet (subway mostly) seven times less dangerous than the streets. PMID:11803819

  5. Accidents due to falls from roof slabs.

    PubMed

    Rudelli, Bruno Alves; Silva, Marcelo Valerio Alabarce da; Akkari, Miguel; Santili, Claudio

    2013-01-01

    CONTEXT AND OBJECTIVE Falls from the roof slabs of houses are accidents of high potential severity that occur in large Brazilian cities and often affect children and adolescents. The aims of this study were to characterize the factors that predispose towards this type of fall involving children and adolescents, quantify the severity of associated lesions and suggest preventive measures. DESIGN AND SETTING Descriptive observational prospective longitudinal study in two hospitals in the metropolitan region of São Paulo. METHODS Data were collected from 29 cases of falls from roof slabs involving children and adolescents between October 2008 and October 2009. RESULTS Cases involving males were more prevalent, accounting for 84%. The predominant age group was schoolchildren (7 to 12 years old; 44%). Leisure activities were most frequently being practiced on the roof slab at the time of the fall (86%), and flying a kite was the most prevalent game (37.9%). In 72% of the cases, the children were unaccompanied by an adult responsible for them. Severe conditions such as multiple trauma and traumatic brain injuries resulted from 79% of the accidents. CONCLUSION Falls from roof slabs are accidents of high potential severity, and preventive measures aimed towards informing parents and guardians about the dangers and risk factors associated with this type of accident are needed, along with physical protective measures, such as low walls around the slab and gates with locks to restrict free access to these places. PMID:23903263

  6. ANS severe accident program overview & planning document

    SciTech Connect

    Taleyarkhan, R.P.

    1995-09-01

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

  7. [Case Report - Really a diving accident?].

    PubMed

    Fichtner, Andreas

    2015-10-01

    A 17 y old male SCUBA diver presents himself for hospital admission after a suspected diving accident. All clinical signs are favouring the initial diagnosis: loss of leg motor function, paresthesia, disturbed vision and headache. What are your further diagnostic and therapeutic steps? Can you proof the initial diagnosis? What differential diagnoses are relevant or even mimicked? PMID:26510103

  8. Accident Prevention: A Workers' Education Manual.

    ERIC Educational Resources Information Center

    International Labour Office, Geneva (Switzerland).

    Devoted to providing industrial workers with a greater knowledge of precautionary measures undertaken and enforced by industries for the protection of workers, this safety education manual contains 14 lessons ranging from "The Problems of Accidents during Work" to "Trade Unions and Workers and Industrial Safety." Fire protection, safety equipment…

  9. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 4 2012-07-01 2011-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  10. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 4 2014-07-01 2013-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  11. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 4 2013-07-01 2013-07-01 false Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  12. Psychological morbidity associated with motor vehicle accidents.

    PubMed

    Blanchard, E B; Hickling, E J; Taylor, A E; Loos, W R; Gerardi, R J

    1994-03-01

    Fifty victims of recent motor vehicle accidents (MVAs), who had sought medical attention after their accidents, were assessed for possible psychological morbidity as a result of the accident. Forty age, gender-matched controls were also assessed with the same instruments. Forty-six percent of the MVA victims met the criteria for current post-traumatic stress disorders (PTSD) as a result of the accident while 20% showed a sub-syndromal version (the reexperiencing symptom cluster plus either the avoidance/numbing cluster or the over-arousal cluster) of PTSD. Although all MVA victims showed some form of driving reluctance, only 1 S met the criteria for driving phobia. Those MVA victims who met the criteria for PTSD or sub-syndromal PTSD were significantly more likely to have experienced previous trauma, other than a serious MVA, and were more likely (P = 0.008) to have previously met the criteria for PTSD as a result of that trauma. Forty-eight percent of MVA victims who met the criteria for current PTSD also met the criteria for current major depression. Significantly more current MVA-PTSDs had suffered previous major depressive episodes. PMID:8192626

  13. RECENT LASER ACCIDENTS AT DEPARTMENT OF ENERGY LABORATORIES

    SciTech Connect

    ODOM, CONNON R.

    2007-02-02

    Recent laser accidents and incidents at research laboratories across the Department of Energy complex are reviewed in this paper. Factors that contributed to the accidents are examined. Conclusions drawn from the accident reports are summarized and compared. Control measures that could have been implemented to prevent the accidents will be summarized and compared. Recommendations for improving laser safety programs are outlined and progress toward achieving them are summarized.

  14. Early dose assessment following severe radiation accidents

    SciTech Connect

    Goans, R.E.; Holloway, E.C.

    1996-06-01

    Prompt and aggressive treatment of victims to high level whole-body gamma exposure has been shown to improve their likelihood of survival. However, in such cases, both the magnitude of the accident and the dosimetry profile(s) of the victim(s) are often not known in detail for days to weeks. Medical intervention could therefore be delayed after a major accident because of uncertainties in the initial dose estimate. A simple dose-prediction algorithm based on lymphocyte kinetics as documented in prior radiation accidents is presented here. This algorithm provides an estimate of marrow dose within the first 12-18 h following an acute whole-body gamma exposure. Early lymphocyte depletion curves post-accident follow a single exponential, L(t) = L{sub o}e{sup -k(D)t}, where L{sub o} is the pre- accident lymphocyte count and k(D) is a rate constant, dependent on the average dose, D. Within the first 12-18 h post-accident, K(D) may be calculated utilizing serial lymphocyte counts. Data from the REAC/TS Accident Registry were used to develop a dose prediction algorithm from 43 gamma exposure cases where both lymphocyte kinetics and dose reconstruction were felt to be reasonably reliable. The relationship D(K) is shown to follow a logistic dose response curve of the form D = a/[1 + (K/b){sup c}] in the range 0 {le} D {le} 15 Gy. The fitting parameters (mean {+-} SD) are found to be a = 21.5 {+-} 5.8 Gy, b = 1.75 {+-} 0.99 d{sup -1}, and c = -0.98 {+-} 0.14, respectively. The coefficient of determination r{sup 2} for the fit is 0.90 with an F-value of 174.7. Dose estimated in this manner is intended to serve only as a first approximation to guide initial medical-management. The treatment regimen may then be modified as needed after more exact dosimetry has become available.

  15. Utilization of accident databases and fuzzy sets to estimate frequency of HazMat transport accidents.

    PubMed

    Qiao, Yuanhua; Keren, Nir; Mannan, M Sam

    2009-08-15

    Risk assessment and management of transportation of hazardous materials (HazMat) require the estimation of accident frequency. This paper presents a methodology to estimate hazardous materials transportation accident frequency by utilizing publicly available databases and expert knowledge. The estimation process addresses route-dependent and route-independent variables. Negative binomial regression is applied to an analysis of the Department of Public Safety (DPS) accident database to derive basic accident frequency as a function of route-dependent variables, while the effects of route-independent variables are modeled by fuzzy logic. The integrated methodology provides the basis for an overall transportation risk analysis, which can be used later to develop a decision support system. PMID:19250750

  16. Magnetorotational iron core collapse

    NASA Technical Reports Server (NTRS)

    Symbalisty, E. M. D.

    1984-01-01

    During its final evolutionary stages, a massive star, as considered in current astrophysical theory, undergoes rapid collapse, thereby triggering a sequence of a catastrophic event which results in a Type II supernova explosion. A remnant neutron star or a black hole is left after the explosion. Stellar collapse occurs, when thermonuclear fusion has consumed the lighter elements present. At this stage, the core consists of iron. Difficulties arise regarding an appropriate model with respect to the core collapse. The present investigation is concerned with the evolution of a Type II supernova core including the effects of rotation and magnetic fields. A simple neutrino model is developed which reproduced the spherically symmetric results of Bowers and Wilson (1982). Several two-dimensional computational models of stellar collapse are studied, taking into account a case in which a 15 solar masses iron core was artificially given rotational and magnetic energy.

  17. Contaminated Sediment Core Profiling

    EPA Science Inventory

    Evaluating the environmental risk of sites containing contaminated sediments often poses major challenges due in part to the absence of detailed information available for a given location. Sediment core profiling is often utilized during preliminary environmental investigations ...

  18. Core assembly storage structure

    DOEpatents

    Jones, Jr., Charles E.; Brunings, Jay E.

    1988-01-01

    A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.

  19. Geophysics: Earth's core problem

    NASA Astrophysics Data System (ADS)

    Dobson, David

    2016-06-01

    Measurements of the electrical resistance and thermal conductivity of iron at extreme pressures and temperatures cast fresh light on controversial numerical simulations of the properties of Earth's outer core. See Letters p.95 & 99

  20. Biospecimen Core Resource - TCGA

    Cancer.gov

    The Cancer Genome Atlas (TCGA) Biospecimen Core Resource centralized laboratory reviews and processes blood and tissue samples and their associated data using optimized standard operating procedures for the entire TCGA Research Network.