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Sample records for loss-of-coolant accident analysis

  1. Analysis of Loss-of-Coolant Accidents in the NBSR

    SciTech Connect

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  2. 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.

  3. Estimating Loss-of-Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models

    SciTech Connect

    S. A. Eide; D. M. Rasmuson; C. L. Atwood

    2008-09-01

    The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.

  4. 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.

  5. 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.

  6. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    SciTech Connect

    Al-Falahi, A.; Haennine, M.; Porkholm, K.

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  7. TRAC large-break loss-of-coolant accident analysis for the AP600 design

    SciTech Connect

    Lime, J.F.; Boyack, B.E.

    1994-02-01

    This report discusses a TRAC model of the Westinghouse AP600 advanced reactor design which has been developed for analyzing large-break loss-of-coolant accident (LBLOCA) transients. A preliminary LBLOCA calculation of a 80% cold-leg break has been performed with TRAC-PF1/MOD2. The 80% break size was calculated by Westinghouse to be the most severe large-break size. The LBLOCA transient was calculated to 92 s. Peak clad temperatures (PCT) were well below the Appendix K limit of 1478 K (2200{degrees}F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code.

  8. Analysis of an AP600 intermediate-size loss-of-coolant accident

    SciTech Connect

    Boyack, B.E.; Lime, J.F.

    1995-04-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  9. Analysis of an AP600 intermediate-size loss-of-coolant accident

    SciTech Connect

    Boyack, B.E.; Lime, J.F.

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  10. 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.

  11. Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

    NASA Astrophysics Data System (ADS)

    Komen, E. M. J.; Koning, H.

    1994-03-01

    This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.

  12. Loss of coolant analysis for the tower shielding reactor 2

    SciTech Connect

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

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

    SciTech Connect

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

    2008-07-31

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

  14. 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.

  15. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    SciTech Connect

    Not Available

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

  16. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    NASA Technical Reports Server (NTRS)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  17. Loss-of-coolant accident experiment at the AVR (Arbeitsgemeinschaft Versuchsreaktor) gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1989-11-01

    Loss of coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into modular gas-cooled reactor designs, loss-of-coolant accident (LOCA) simulation tests were conducted with the 15-MW(electric), 46-MW(thermal), pebble-bed, high-temperature Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany (FRG). This is the only nuclear power plant ever to have been intentionally subjected to LOCa conditions. Oak Ridge National Laboratory participation in the preparation and conduct of the tests was carried out within the U.S./FRG Agreement for Cooperation in Gas-Cooled Reactor Development.

  18. 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.

  19. Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Nakamura, M.; Tobita, K.; Someya, Y.; Utoh, H.; Sakamoto, Y.; Gulden, W.

    2015-11-01

    Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

  20. 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.

  1. 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.

  2. APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.

  3. Emergency cooling analysis for the loss of coolant malfunction

    NASA Technical Reports Server (NTRS)

    Peoples, J. A.

    1972-01-01

    This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.

  4. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

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

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  5. 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.

  6. TRAC loss-of-coolant accident analyses of the Savannah River production reactors

    SciTech Connect

    Lime, J.F.; Motley, F.E. )

    1990-06-01

    TRAC loss-of-coolant accident (LOCA) analyses were performed as part of the independent safety review of the US Department of Energy's Savannah River (SR) production reactors. The double-ended guillotine break in a coolant loop is a design-basis LOCA for the SR reactors. Three break locations were analyzed to determine the worst break location: (1) at the pump-suction flange; (2) at the pump discharge flange; or (3) at the plenum inlet. The plenum-inlet break was shown to be the most severe in terms of minimum flow delivered to each fuel assembly in the reactor core.

  7. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  8. Sensitivity studies of loss-of-coolant accidents in the Savannah River production reactors

    SciTech Connect

    Edwards, J.N.; Motley, F.E.; Morgan, M.M.; Knight, T.D.; Fischer, S.R. )

    1990-01-01

    Loss-of-coolant accident (LOCA) analyses were completed using the Transient Reactor Analysis Code (TRAC) to support the U.S. Department of Energy efforts to restart the production reactors located at the Savannah River Site. The break location and pump operation after the LOCA were the parameters varied for these sensitivity studies. Three location of double-ended guillotine break were studied: plenum inlet, pump suction, and pump discharge. Three pump operation scenarios were also studied: continued operation of both ac and dc pumps, tripping of the ac motor at 2 s after the LOCA, and tripping of the ac motor at 200 s after the LOCA. The production reactors use low pressure and temperature heavy water as the process fluid. The reactor has a moderator tank that contains the fuel channels. Above the moderator tank is an upper plenum that distributes the heavy water to each fuel assembly. The heavy water flows down through the fuel channels and into the moderator tank. From the tank, the water is pumped back to the upper plenum through six loops. Each loop contains a pump and two heat exchangers. Four of the loops have an emergency core coolant system (ECCS) connection. This TRAC model has been benchmarked extensively against data taken in the actual reactors or in prototypical models of the components of the reactors. The calculations were completed using a version of TRAC-PF1/MOD 2 that was updated to include heavy water properties and other changes that are specific to the production reactors.

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

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1990-01-01

    A landmark safety test has been conducted at the AVR-reactor, a high-temperature gas-cooled reactor (HTGR) in the Federal Republic of Germany owned by the Arbeitsgemeinschaft Versuchsreaktor, AVR in Juelich. The 46-MW(t), 15-MW(e) AVR reactor was subjected to a simulated loss-of-coolant accident (LOCA), a very severe occurrence in which the coolant escapes from the reactor core and no emergency system provides coolant flow to the core. The test, which demonstrated the inherently safe response of this reactor to a LOCA, marked the first time ever that a reactor has been intentionally subjected to loss-of-coolant conditions without emergency cooling. Oak Ridge National Laboratory (ORNL) and General Atomics participated in the test by working with AVR staff by jointly performing the analyses needed to obtain the license to conduct the test and by performing post test analyses. This participation was carried out under the cooperative AVR Subprogram which is conducted within the US/FRG Agreement for Cooperation in Gas-Cooled Reactor Development. 7 figs.

  10. Transient Analysis for Evaluating the Potential Boiling in the High Elevation Emergency Cooling Units of PWR Following a Hypothetical Loss of Coolant Accident (LOCA) and Subsequent Water Hammer Due to Pump Restart

    SciTech Connect

    Husaini, S. Mahmood; Qashu, Riyad K.

    2004-07-01

    The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is restored, the Component Cooling Water (CCW) pumps restart causing water hammer to occur due to cavity closure. Recently EPRI (Electric Power Research Institute) performed a research study that recommended a methodology to mitigate the water hammer due to cavity closure. The EPRI methodology allows for the cushioning effects of hot steam and released air, which is not considered in the conventional water column separation analysis. The EPRI study was limited in scope to the evaluation of water hammer only and did not provide any guidance for evaluating the occurrence of boiling and the extent of voiding in the ECU piping. This paper presents a complete methodology based on first principles to evaluate the onset of boiling. Also, presented is a methodology for evaluating the extent of voiding and the water hammer resulting from cavity closure by using an existing generalized computer program that is based on the Method of Characteristics. The EPRI methodology is then used to mitigate the predicted water hammer. Thus it overcomes the inherent complications and difficulties involved in performing hand calculations for water hammer. The heat transfer analysis provides an alternative to the use of very cumbersome modeling in using CFD (computational fluid dynamics) based computer programs. (authors)

  11. Cobalt-60 simulation of LOCA (loss of coolant accident) radiation effects

    SciTech Connect

    Buckalew, W.H.

    1989-07-01

    The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs.

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

    SciTech Connect

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-07-01

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

  13. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    SciTech Connect

    Kao, S.P.; Chang, S.K.; Huang, H.C.

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

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

    SciTech Connect

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

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

  15. 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.

  16. Loads on steam generator tubes during simulated loss-of-coolant accident conditions. Final report. [PWR

    SciTech Connect

    Guerrero, H.N.; Hiestand, J.W.; Rossano, F.V.; Shah, P.K.; Thakkar, J.G.

    1982-11-01

    This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditions in an operational nuclear steam generator. The loop consisted of 5 full size double 90/sup 0/ bend tubes and steam generator plena, a pressurizer, a reactor resistance simulator, a heater, a pump, and associated pipes and valves to complete the system. The tubes used were of typical length and the same outside diameter as those used in C-E steam generators. Prototypical supports were provided for the bundle of 5 tubes. Cold leg guillotine breaks were simulated using quick opening valve and rupture disks. Break opening times ranged from less than 1 msec to as much as 67 milliseconds. The loop instrumentation was designed to measure the transient pressure history at various locations and monitor the structural response of the tube to the LOCA hydrodynamic loading. A series of blowdown tests was performed for different operating and boundary conditions. Analytically predicted transient pressure histories and the differential pressure history across the tube span were compared with the experimental data.

  17. Probabilistic assessment of decoupling loss-of-coolant accident and earthquake in nuclear power plant design

    SciTech Connect

    Lu, S.C.; Harris, D.O.

    1981-01-01

    This paper describes a research project conducted at Lawrence Livermore National Laboratory to establish a technical basis for reassessing the requirement of combining large loss-of-coolant-accident (LOCA) and earthquake loads in nuclear power plant design. A large LOCA is defined herein as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressureized water reactor (PWR). A systematic probability approach has been employed to estimate the probability of a large LOCA directly and indirectly induced by earthquakes. The probability of a LOCA directly induced by earthquakes was assessed by a numerical simulation of pipe rupture of a reactor coolant system. The simulation employed a deterministic fracture mechanics model which dictates the fatigue growth of pre-existing cracks in the pipe. The simulation accounts for the stochastic nature of input elements such as the initial crack size distribution, the crack occurrence rate, crack and leak detection probabilities as functions of crack size, plant transient occurrence rates, the seismic hazard, stress histories, and crack growth model parameters. Effects on final results due to variation an uncertainty of input elements were assessed by a limited sensitivity study. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the orer of 10/sup -12/). The probability of a leak was found to be several orders of magnitudes greater than that of a complete break.

  18. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    SciTech Connect

    Nelson, C.F.; Gauthier, G.; Carlin, F.

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  19. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    SciTech Connect

    Nelson, C.F.

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  20. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  1. 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.

  2. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  3. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  4. Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions

    SciTech Connect

    Chung, H.M.; Neimark, L.A.; Kassner, T.F.

    1996-10-01

    Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.

  5. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  6. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    SciTech Connect

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  7. 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)

  8. Assessment of a large break loss of coolant accident scenario requiring operator action to initiate safety injection

    SciTech Connect

    Grendys, R.C.; Nissley, M.E.; Baker, D.C.

    1996-11-01

    As part of the licensing basis for a nuclear power plant, the acceptability of the Emergency Core Cooling Systems (ECCS) following a postulated Loss-of-Coolant Accident (LOCA) as described in the Code of Federal Regulations (CFR), Title 10, Chapter 1, Part 50.46, must be verified. The LOCA analysis is performed with an acceptable ECCS Evaluation Model and results must show compliance with the 10 CFR 50.46 acceptance criteria. Westinghouse Electric Corporation performs Large and Small Break LOCA and LOCA-related analyses to support the licensing basis of various nuclear power plants and also performs evaluations against the licensing basis analyses as required. Occasionally, the need arises for the holder of an operating license of a nuclear power plant to submit a Licensee Event Report (LER) to the US Nuclear Regulatory Commission (USNRC) for any event of the type described in the Code of Federal Regulations, Title 10, Chapter 1, Part 50.73. To support the LER, a Justification for Past Operation (JPO) may be performed to assess the safety consequences and implications of the event based on previous operating conditions. This paper describes the work performed for the Large Break LOCA to assess the impact of an event discovered by Florida Power and Light and reported in LER-94-005-02. For this event, it was determined that under certain circumstances, operator action would have been required to initiate safety injection (SI), thus challenging the acceptability of the ECCS. This event was specifically addressed for the Large Break LOCA by using an advanced thermal hydraulic analysis methodology with realistic input assumptions.

  9. Experimental investigation on the chemical precipitation generation under the loss of coolant accident of nuclear power plants

    SciTech Connect

    Kim, C. H.; Sung, J. J.; Chung, Y. W.

    2012-07-01

    The PWR containment buildings are designed to facilitate core cooling in the event of a Loss of Coolant Accident (LOCA). The cooling process requires water discharged from the break and containment spray to be collected in a sump for recirculation. The containment sump contains screens to protect the components of the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) from debris. Since the containment materials may dissolve or corrode when exposed to the reactor coolant and spray solutions, various chemical precipitations can be generated in a post-LOCA environment. These chemical precipitations may become another source of debris loading to be considered in sump screen performance and downstream effects. In this study, new experimental methodology to predict the type and quantity of chemical precipitations has been developed. To generate the plant-specific chemical precipitation in a post-LOCA environment, the plant specific chemical condition of the recirculation sump during post-LOCA is simulated with the experimental reactor for the chemical effect. The plant-specific containment materials are used in the present experiment such as glass fibers, concrete blocks, aluminum specimens, and chemical reagent - boric acid, spray additives or buffering chemicals (sodium hydroxide, Tri-Sodium Phosphate (TSP), or others). The inside temperature of the reactor is controlled to simulate the plant-specific temperature profile of the recirculation sump. The total amount of aluminum released from aluminum specimens is evaluated by ICP-AES analysis to determine the amount of AlOOH and NaAlSi{sub 3}O{sub 8} which induce very adverse effect on the head loss across the sump screens. The amount of these precipitations generated in the present experimental study is compared with the results of WCAP-16530-NP-A. (authors)

  10. Dose to man from a hypothetical loss-of-coolant accident at the Rancho Seco Nuclear Power Plant

    SciTech Connect

    Peterson, K.R.; Greenly, G.D.

    1981-02-01

    At the request of the Sacramento Municipal Utilities District, we used our computer codes, MATHEW and ADPIC, to assess the environmental impact of a loss-of-coolant accident at the Rancho Seco Nuclear Power Plant, about 40 kilometres southeast of Sacramento, California. Meteorological input was selected so that the effluent released by the accident would be transported over the Sacramento metropolitan area. With the release rates provided by the Sacramento Municipal Utilities District, we calculated the largest total dose for a 24-hour release as 70 rem about one kilometre northwest of the reactor. The largest total dose in the Sacramento metropolitan area is 780 millirem. Both doses are from iodine-131, via the forage-cow-milk pathway to an infant's thyroid. The largest dose near the nuclear plant can be minimized by replacing contaminated milk and by giving the cows dry feed. To our knowledge, there are no milk cows within the Sacramento metropolitan area.

  11. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect

    Simpson, D.B.; Greene, S.R.

    1990-01-01

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

  12. EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

    1981-04-01

    Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

  13. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    SciTech Connect

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  14. LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor: Appendix A-4

    SciTech Connect

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.

  15. 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.

  16. Heat transfer processes during intermediate and large break loss-of-coolant accidents (LOCAs)

    SciTech Connect

    Vojtek, I

    1986-09-01

    The general purpose of this project was the investigation of the heat transfer regimes during the high pressure portion of blowdown. The main attention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimental results of the 25-rod bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karlstein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The computer code BRUDI-VA was used for the calculation of local values of important thermohydraulic parameters in the bundle.

  17. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    SciTech Connect

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

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  18. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    SciTech Connect

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  19. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    NASA Astrophysics Data System (ADS)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  20. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  1. Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

    NASA Astrophysics Data System (ADS)

    Ha, Taesung

    A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

  2. A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility.

    SciTech Connect

    Wachs, D. M.

    1998-11-04

    Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS.

  3. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

    SciTech Connect

    Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

    1981-09-01

    A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

  4. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R. ); Choi, C.J. )

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  5. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  6. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

  7. A thermodynamic model for noble metal alloy inclusions in nuclear fuel rods and application to the study of loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Kaye, Matthew Haigh

    Metal alloy inclusions comprised of Mo, Pd, Rh, Ru, and Tc (the so-called "noble" metals) develop in CANDU fuel pellets as a result of fission. The thermochemical behaviour of this alloy system during severe accident conditions is of interest in connection with computations of loss of volatile compounds of these elements by reaction with steam-hydrogen gas mixtures that develop in the system as a result of water reacting with the Zircalloy cladding. This treatment focuses on the development of thermodynamic models for the Mo-Pd-Rh-Ru-Tc quinary system. A reasonable prediction was made by modelling the ten binary phase diagrams, five of these evaluations being original to this work. This process provides a complete treatment for the five solution phases (vapour, liquid, bcc-solid, fcc-solid, and cph-solid) in this alloy system, as well as self-consistent Gibbs energies of formation for the Mo 5Ru3 intermetallic phase, and two intermediate phases in the Mo-Tc system. The resulting collection of properties, when treated by Gibbs energy minimization, permits phase equilibria to be computed for specified temperatures and compositions. Experimental work in support of this treatment has been performed. Measurements of the solidus and liquidus temperatures for Pd-Rh alloys were made using differential thermal analysis. These measurements confirm that the liquid solution exhibits positive deviation from Raoult's law. Experimental work as a visiting research engineer at AECL (Chalk River) was performed using a custom developed Knudsen cell/mass spectrometer. The Pd partial pressure was measured above multi-component alloys of known composition over a range of temperatures. These are correlated to predicted activities of Pd from the developed thermodynamic model in the multi-component alloy. The thermodynamic treatment developed for the noble metal alloy inclusions has been combined with considerable other data and applied to selected loss-of-coolant-accident scenarios to

  8. Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

    SciTech Connect

    Loomis, G G; Streit, J E

    1985-11-01

    Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature (15.6 MPa (2262 psia) system pressure; 37 K (67F) core differential temperature; 595 K(610F) hot leg fluid temperature). From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs.

  9. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    SciTech Connect

    Ghan, L.S.; Ortiz, M.G.

    1991-12-31

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B&W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission`s (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B&W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions.

  10. 77 FR 19740 - Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-02

    ... accident. II. Further Information DG-1234 was published in the Federal Register on July 15 2010, (75 FR... atmosphere cleanup systems. RG 1.82 provides guidelines for evaluating the adequacy and the availability of... containment atmosphere cleanup systems. RG 1.82 provides guidelines for evaluating the adequacy and...

  11. Multiloop integral system test (MIST): Test Group 31, SBLOCA (small-break loss-of-coolant accident) with varied boundary conditions

    SciTech Connect

    Gloudemans, J.R. . Nuclear Power Div.)

    1989-07-01

    The multiloop integral system test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox-designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock and Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral system facilities to address the thermal-hydraulic SBLOCA questions. MIST and two other supporting facilities were specifically designed and constructed for this program, and an existing facility --- the once-through integral system (OTIS) --- was also used. Data from MIST and the other facilities will be used to benchmark the adequacy of system codes, such as RELAP-5 and TRAC, for predicting abnormal plant transients. The MIST program is reported in 11 volumes. The program is summarized in Volume 1; Volumes 2 through 8 describe groups of tests by test type; Volume 9 presents inter-group comparisons; Volume 10 provides comparisons between the calculations of RELAP5 MOD2 and MIST observations, and Volume 11 presents the later, Phase 4 tests. This is Volume 3 pertaining to Test Group 31, Boundary Conditions Variations. The specifications, conduct, observations, and results of these tests are described. 8 refs., 328 figs., 15 tabs.

  12. Matrix database of EPRI experimental projects on small-break loss-of-coolant accidents: Final report

    SciTech Connect

    House, R.K.

    1989-08-01

    The large break LOCA had been studied extensively for many years and substantial analytical methods had been developed and verified for evaluation of this DBA. The nature of the physical processes which occur during a SBLOCA are significantly different than those which occur in a DBA LOCA, especially for a pressurized water reactor (PWR). As a result, only a portion of the methods developed for large break LOCA evaluation were applicable to SBLOCA, and a new list of research priorities arose out of questions raised during the evaluation of TMI-2. The nuclear safety research and development program at EPRI responded to the industry's need for improved understanding of SBLOCA. Projects to investigate SBLOCA that were in place prior to the TMI-2 accident were augmented and new projects were implemented. The overall cope of EPRI SBLOCA projects include both analytical development and experimental work. This report focuses on the experimental work sponsored by EPRI to support improved understanding of the SBLOCA. All of the EPRI experimental projects which relate to SBLOCA, both prior to and following TMI-2, were reviewed and the results summarized in an attempt to provide an overall matrix oriented view of the EPRI SBLOCA experimental database. 5 tabs.

  13. Blanket Module Boil-Off Times during a Loss-of-Coolant Accident - Case 0: with Beam Shutdown only

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document LBLOCA analyses for the Accelerator Production of Tritium primary blanket Heat Removal system. This report documents the analysis results of a LBLOCA where the accelerator beam is shut off without primary pump trips and neither the RHR nor the cavity flood systems operation.

  14. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  15. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  16. APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  17. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  18. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    SciTech Connect

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A.; Voeroess, L.; Techy, Z.; Katona, T.

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  19. Multi-Pin Studies of the Effect of Changes in PWR Fuel Design on Clad Ballooning and Flow Blockage in a Large-Break Loss-Of Coolant Accident

    SciTech Connect

    Jones, J.R.; Trow, M.

    2007-07-01

    Fuel pins can credibly balloon to reach very high diametric strains under temperature transients typical of a PWR Loss-of coolant Accident (LOCA), but experiments show that these balloons are sufficiently misaligned axially to prevent total blockage of the flow. Most of the relevant experiments were performed in the 1980's and therefore were principally carried out on the various forms of Zircaloy 4 cladding available at the time. Much of the fuel used was either fresh or of modest burnup compared to the discharge irradiations achievable today. Since then, single pin experiments have been carried out with new cladding material and (to a limited extent) with high-burnup fuel. However, there is a need to interpret the performance of this fuel in the context of the wider body of evidence. A model of the development of flow blockages has been implemented using multiple instances of the fuel pin code MABEL interfaced to a sub-channel coolant flow code. The effect of a change in cladding material from Zircaloy to a 1% niobium alloy has been examined. The assessment concluded that the proposed replacement alloy is more creep hard at high temperature and therefore is expected to fail slightly later in the transient. The new cladding achieved a generally lower diametric strain at failure under the particular conditions of the fault. (authors)

  20. APT Blanket System Loss-of-Coolant Analysis Based on Initial Conceptual Design - Case 2: External HR Break HR Break at Pump Outlet with Pump Trip

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  1. TRAC-PF1/MOD1 calculations and data comparisons for MIST (Multi-Loop Integral System Test) small-break loss-of-coolant accidents with scaled 10 cm/sup 2/ and 50 cm/sup 2/ breaks

    SciTech Connect

    Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

    1987-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents (SBLOCAs), loss of feedwater and other transients in Babcock and Wilcox (B and W) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 x 4 (2 hot legs and steam generators, 4 cold legs and reactor-coolant pumps) representation of lowered-loop reactor systems of the B and W design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at Stanford Research Institute. The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are underway at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment have been completed for two transients run in the MIST facility. These are the MIST nominal test. Test 3109AA, a scaled 10 cm/sup 2/ SBLOCA and Test 320201, a scaled 50 cm/sup 2/ SBLOCA. Only MIST assessment results are presented in this paper.

  2. Nuclear reactor loss of coolant protection system

    SciTech Connect

    Loose, R.A.

    1986-03-18

    A pressurized water reactor system is described of a nuclear power plant having a water storage tank for providing emergency coolant water and means provided external to the containment vessel, for use in the event of a primary loss of coolant situation, to circulate emergency water as a coolant by withdrawal through a wall of the containment vessel and return the same back through the wall of the containment vessel and passing the water through a heat exchange means prior to use as a coolant for the reactor core. The improvement described here consists of: an enslosure, the interior of which is sealed to the atmosphere, positioned adjacent to and exterior of a wall of the containment vessel; an inlet conduit, enclosed within a sealed outer casing, communicating between the interior of the containment vessel and the interior of the enclosure; an exhaust conduit, enclosed within a sealed outer casing, communicating between the interior of the enclosure and the interior of the containment vessel; a rupture disc on the inlet conduit within the enclosure, such that failure of the exhaust conduit within the enclosure will produce an increase of the pressure within the enclosure and above a predetermined pressure will fracture the rupture disc, and will circulate the coolant within the enclosure; and means within the interior of the enclosure for pumping coolant from the interior of the containment vessel through the inlet conduit, and back to the interior of the containment vessel through the exhaust conduit; whereby if either of the conduits should fail, coolant will be collected within the enclosure and sealed to the atmosphere.

  3. Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

    SciTech Connect

    Varacalle, Jr., D. J.; MacDonald, P. E.; Shiozawa, S.; Driskell, W. E.

    1980-01-01

    Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and system pressure measurements with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements.

  4. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  5. 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.

  6. Progress in accident analysis of the HYLIFE-II inertial fusion energy power plant design

    SciTech Connect

    Reyes, S; Latkowski, J F; Gomez del Rio, J; Sanz, J

    2000-10-11

    The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li{sub 2}BeF{sub 4}) was lost at the beginning of the accident. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that his dose was dominated by the tritium released during the accident.

  7. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

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

    SciTech Connect

    Peng, W.; Yang, X. Y.; Yu, S. Y.; Wang, J.

    2012-07-01

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

  9. Solar receiver protection means and method for loss of coolant flow

    DOEpatents

    Glasgow, L.E.

    1980-11-24

    An apparatus and method are disclosed for preventing a solar receiver utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver by a plurality of reflectors which rotate so that they direct solar energy to the receiver as the earth rotates. The apparatus disclosed includes a first storage tank for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank having an inlet through which the coolant can enter. The first and second storage tanks are in fluid communication with each other through the solar receiver. The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank through the solar receiver and into the second storage tank. Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks will be sufficient to maintain the coolant in the receiver below a predetermined upper temperature until the solar reflectors become defocused with respect to the solar receiver due to the earth's rotation.

  10. Solar receiver protection means and method for loss of coolant flow

    DOEpatents

    Glasgow, Lyle E.

    1983-01-01

    An apparatus and method for preventing a solar receiver (12) utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver (12) by a plurality of reflectors (16) which rotate so that they direct solar energy to the receiver (12) as the earth rotates. The apparatus disclosed includes a first storage tank (30) for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank (30) includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank (34) for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank (34) having an inlet through which the coolant can enter. The first and second storage tanks (30) and (34) are in fluid communication with each other through the solar receiver (12). The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank (30) through the solar receiver (12) and into the second storage tank (34). Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors (16) stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks (30) and (34) will be sufficient to maintain the coolant in the receiver (12) below a predetermined upper temperature until the solar reflectors (16) become defocused with respect to the solar receiver (12) due to the earth's rotation.

  11. Development of thermal-hydraulic analysis capabilities for Oyster creek

    SciTech Connect

    Lee, R.B.

    1987-01-01

    GPU Nuclear (GPUN) has been involved in developing analytical methodologies for Oyster Creek plant thermal-hydraulic response simulation for approx. 15 yr. Plant-system-related transient analysis is being accomplished via RETRAN02 MOD4 and loss-of-coolant accident (LOCA) analysis by SAFER-CORECOOL. This paper reviews the developmental process and lessons learned through this process.

  12. 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.

  13. 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.

  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. Neutron Radiography and Tomography Investigations of the Secondary Hydriding of Zircaloy-4 during Simulated Loss of Coolant Nuclear Accidents

    NASA Astrophysics Data System (ADS)

    Grosse, Mirco K.; Stuckert, Juri; Steinbrück, Martin; Kaestner, Anders P.; Hartmann, Stefan

    In the framework of the post-test examinations of the large-scale LOCA simulation tests at the fuel rod bundle scale, the hydrogen distributions in specimens prepared from the QUENCH-L0 and -L1 tests were studied by means of neutron radiography and tomography. In order to determine quantitative hydrogen concentrations, both, neutron radiography and tomography were calibrated using cladding tube segments with known hydrogen concentrations. The linear dependence of the total macroscopic neutron cross section with the H/Zr atomic ratio was determined for both methods. The hydrogen distributions in samples prepared from the two tests differ significantly as a first glance to the results obtained for the QUENCH-L1 shows. Whereas clearly visible hydrogen bands were found in samples of the QUENCH-L0 test with a time between burst and quenching of more than 70 s; in some specimens prepared from the QUENCH-L1 test only blurred bands could be detected. The reasons for these different behaviors can be the different times between reaching the temperature maxima and the quenching, as well as bending of the QUENCH-L1 bundle. In the QUENCH-L0 test the bundle was quenched immediately after reaching the maximal temperature. In QUENCH-L1 the hydrogen had about 130 s to diffuse and reach more homogeneous distributions without clear contrasts between the hydrogen bands and the neighboring regions in the neutron images.

  17. 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)

  18. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    NASA Astrophysics Data System (ADS)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  19. Validation and verification of RELAP5 for Advanced Neutron Source accident analysis: Part I, comparisons to ANSDM and PRSDYN codes

    SciTech Connect

    Chen, N.C.J.; Ibn-Khayat, M.; March-Leuba, J.A.; Wendel, M.W.

    1993-12-01

    As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety analysis. Both the ANSDM and PRSDYN models use a simplified single-phase equation set to predict transient thermal-hydraulic performance. Brief descriptions of each of the codes, models, and model limitations were included. Even though comparisons were limited to single-phase conditions, a broad spectrum of accidents was benchmarked: a small loss-of-coolant-accident (LOCA), a large LOCA, a station blackout, and a reactivity insertion accident. The overall conclusion is that the three models yield similar results if the input parameters are the same. However, ANSDM does not capture pressure wave propagation through the coolant system. This difference is significant in very rapid pipe break events. Recommendations are provided for further model improvements.

  20. AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

    SciTech Connect

    Knight, T.D.

    1996-07-01

    The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA.

  1. 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.

  2. Seismic fragility analysis of buried steel piping at P, L, and K reactors

    SciTech Connect

    Wingo, H.E.

    1989-10-01

    Analysis of seismic strength of buried cooling water piping in reactor areas is necessary to evaluate the risk of reactor operation because seismic events could damage these buried pipes and cause loss of coolant accidents. This report documents analysis of the ability of this piping to withstand the combined effects of the propagation of seismic waves, the possibility that the piping may not behave in a completely ductile fashion, and the distortions caused by relative displacements of structures connected to the piping.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. BESAFE II: Accident safety analysis code for MFE reactor designs

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

  8. 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.

  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. 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.

  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. 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.

  13. 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.

  14. 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)

  15. 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

  16. Analysis of thermosyphoning in a steam generator model

    SciTech Connect

    Martinez, M.J.; Martinez, G.M.

    1991-10-01

    An analysis of thermosyphoning in a stream generator model is presented. The model considers the transient development of buoyancy-driven steam flow in the steam generator tubing, secondary side heat transfer and an inlet plenum mixing model. Numerical solutions are obtained for conditions intended to simulate the natural circulation phenomena in a 3-Loop pressurized water reactor in a loss-of-coolant accident scenario. The relation between the circulation rate and the heating rate is determined. The sensitivity of the model to various key parameters is examined. 16 refs.

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

    SciTech Connect

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

    2000-03-23

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

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

    SciTech Connect

    CROWE, R.D.

    1999-09-09

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

  19. Canister storage building design basis accident analysis documentation

    SciTech Connect

    KOPELIC, S.D.

    1999-02-25

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

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

    SciTech Connect

    PIEPHO, M.G.

    1999-10-20

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

  1. User's guide for the PWR LOCA analysis capability of the WRAP-EM system

    SciTech Connect

    Beranek, F; Gregory, M V

    1980-02-01

    The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input specifications for modules in the WRAP-EM system are presented in this document along with the JOSHUA input templates. This document, along with the WRAP user's guide, provides a step-by-step procedure for setting up a PWR data base for the WRAP-EM system. 12 refs.

  2. Analysis of containment venting at the Peach Bottom Atomic Power Station

    SciTech Connect

    Hanson, D.J.; Blackman, H.S.; Nelson, W.R.; Wright, R.E.; Leonard, M.T.; DiSalvo, R.

    1986-10-24

    An analysis of the extent to which containment venting would be effective in preventing or mitigating the consequences of severe accidents has been completed for the Peach Bottom Atomic Power Station, Units 2 and 3 (BWR-4s with Mark I containments). The analysis indicates that the effectiveness of venting in preventing containment overpressurization highly depends on the sequence of the severe accident. Containment venting can be effective for several classes of sequences, including transients with failure of long-term decay heat removal and loss-of-coolant accidents with breaks inside the containment. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with three severe accident sequences currently identified as important risk contributors at Peach Bottom. Means of improving the potential for risk reduction is identified, but their influence on risk is not analyzed.

  3. 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).

  4. [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

  5. 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.

  6. 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.

  7. Development of Database for Accident Analysis in Indian Mines

    NASA Astrophysics Data System (ADS)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

    Mining is a hazardous industry and high accident rates associated with underground mining is a cause of deep concern. Technological developments notwithstanding, rate of fatal accidents and reportable incidents have not shown corresponding levels of decline. This paper argues that adoption of appropriate safety standards by both mine management and the government may result in appreciable reduction in accident frequency. This can be achieved by using the technology in improving the working conditions, sensitising workers and managers about causes and prevention of accidents. Inputs required for a detailed analysis of an accident include information on location, time, type, cost of accident, victim, nature of injury, personal and environmental factors etc. Such information can be generated from data available in the standard coded accident report form. This paper presents a web based application for accident analysis in Indian mines during 2001-2013. An accident database (SafeStat) prototype based on Intranet of the TCP/IP agreement, as developed by the authors, is also discussed.

  8. NASA's Accident Precursor Analysis Process and the International Space Station

    NASA Technical Reports Server (NTRS)

    Groen, Frank; Lutomski, Michael

    2010-01-01

    This viewgraph presentation reviews the implementation of Accident Precursor Analysis (APA), as well as the evaluation of In-Flight Investigations (IFI) and Problem Reporting and Corrective Action (PRACA) data for the identification of unrecognized accident potentials on the International Space Station.

  9. An analysis of pilot error-related aircraft accidents

    NASA Technical Reports Server (NTRS)

    Kowalsky, N. B.; Masters, R. L.; Stone, R. B.; Babcock, G. L.; Rypka, E. W.

    1974-01-01

    A multidisciplinary team approach to pilot error-related U.S. air carrier jet aircraft accident investigation records successfully reclaimed hidden human error information not shown in statistical studies. New analytic techniques were developed and applied to the data to discover and identify multiple elements of commonality and shared characteristics within this group of accidents. Three techniques of analysis were used: Critical element analysis, which demonstrated the importance of a subjective qualitative approach to raw accident data and surfaced information heretofore unavailable. Cluster analysis, which was an exploratory research tool that will lead to increased understanding and improved organization of facts, the discovery of new meaning in large data sets, and the generation of explanatory hypotheses. Pattern recognition, by which accidents can be categorized by pattern conformity after critical element identification by cluster analysis.

  10. An analysis of pileup accidents in highway systems

    NASA Astrophysics Data System (ADS)

    Chang, Jau-Yang; Lai, Wun-Cing

    2016-02-01

    Pileup accident is a multi-vehicle collision occurring in the lane and producing by successive following vehicles. It is a special collision on highway. The probability of the occurrence of pileup accident is lower than that of the other accidents in highway systems. However, the pileup accident leads to injuries and damages which are often serious. In this paper, we analyze the occurrence of pileup accidents by considering the three types of dangerous collisions in highway systems. We evaluate those corresponding to rear-end collision, lane-changing collision, and double lane-changing collision. We simulate four road driving strategies to investigate the relationships between different vehicle collisions and pileup accidents. In accordance with the simulation and analysis, it is shown that the double lane-changing collisions result in an increase of the occurrence of pileup accidents. Additionally, we found that the probability of the occurrence of pileup accidents can be reduced when the speeds of vehicles are suitably constrained in highway systems.

  11. 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.

  12. Full-length fuel rod behavior under severe accident conditions

    SciTech Connect

    Lombardo, N J; Lanning, D D; Panisko, F E

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

  13. Systemic accident analysis: examining the gap between research and practice.

    PubMed

    Underwood, Peter; Waterson, Patrick

    2013-06-01

    The systems approach is arguably the dominant concept within accident analysis research. Viewing accidents as a result of uncontrolled system interactions, it forms the theoretical basis of various systemic accident analysis (SAA) models and methods. Despite the proposed benefits of SAA, such as an improved description of accident causation, evidence within the scientific literature suggests that these techniques are not being used in practice and that a research-practice gap exists. The aim of this study was to explore the issues stemming from research and practice which could hinder the awareness, adoption and usage of SAA. To achieve this, semi-structured interviews were conducted with 42 safety experts from ten countries and a variety of industries, including rail, aviation and maritime. This study suggests that the research-practice gap should be closed and efforts to bridge the gap should focus on ensuring that systemic methods meet the needs of practitioners and improving the communication of SAA research. PMID:23542136

  14. OFFSITE RADIOLOGICAL CONSEQUENCE ANALYSIS FOR THE BOUNDING FLAMMABLE GAS ACCIDENT

    SciTech Connect

    KRIPPS, L.J.

    2005-02-18

    This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a SST. The calculation applies reasonably conservative input parameters in accordance with guidance in DOE-STD-3009, Appendix A. The purpose of this analysis is to calculate the offsite radiological consequence of the bounding flammable gas accident. DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', requires the formal quantification of a limited subset of accidents representing a complete set of bounding conditions. The results of these analyses are then evaluated to determine if they challenge the DOE-STD-3009-94, Appendix A, ''Evaluation Guideline,'' of 25 rem total effective dose equivalent in order to identify and evaluate safety-class structures, systems, and components. The bounding flammable gas accident is a detonation in a single-shell tank (SST). A detonation versus a deflagration was selected for analysis because the faster flame speed of a detonation can potentially result in a larger release of respirable material. A detonation in an SST versus a double-shell tank (DST) was selected as the bounding accident because the estimated respirable release masses are the same and because the doses per unit quantity of waste inhaled are greater for SSTs than for DSTs. Appendix A contains a DST analysis for comparison purposes.

  15. Analysis of tritium mission FMEF/FAA fuel handling accidents

    SciTech Connect

    Van Keuren, J.C.

    1997-11-18

    The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

  16. Corporate cost of occupational accidents: an activity-based analysis.

    PubMed

    Rikhardsson, Pall M; Impgaard, Martin

    2004-03-01

    The systematic accident cost analysis (SACA) project was carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. Its focused on developing and testing a method for evaluating occupational costs of companies for use by occupational health and safety professionals. The method was tested in nine Danish companies within three different industry sectors and the costs of 27 selected occupational accidents in these companies were calculated. One of the main conclusions is that the SACA method could be used in all of the companies without revisions. The evaluation of accident cost showed that 2/3 of the costs of occupational accidents are visible in the Danish corporate accounting systems reviewed while 1/3 is hidden from management view. The highest cost of occupational accidents for a company with 3.600 employees was estimated to approximately US$ 682.000. The paper includes an introduction regarding accident cost analysis in companies, a presentation of the SACA project methodology and the SACA method itself, a short overview of some of the results of the SACA project and a conclusion. Further information about the project is available at http://www.asb.dk/saca. PMID:14642872

  17. Historical analysis of US pipeline accidents triggered by natural hazards

    NASA Astrophysics Data System (ADS)

    Girgin, Serkan; Krausmann, Elisabeth

    2015-04-01

    Natural hazards, such as earthquakes, floods, landslides, or lightning, can initiate accidents in oil and gas pipelines with potentially major consequences on the population or the environment due to toxic releases, fires and explosions. Accidents of this type are also referred to as Natech events. Many major accidents highlight the risk associated with natural-hazard impact on pipelines transporting dangerous substances. For instance, in the USA in 1994, flooding of the San Jacinto River caused the rupture of 8 and the undermining of 29 pipelines by the floodwaters. About 5.5 million litres of petroleum and related products were spilled into the river and ignited. As a results, 547 people were injured and significant environmental damage occurred. Post-incident analysis is a valuable tool for better understanding the causes, dynamics and impacts of pipeline Natech accidents in support of future accident prevention and mitigation. Therefore, data on onshore hazardous-liquid pipeline accidents collected by the US Pipeline and Hazardous Materials Safety Administration (PHMSA) was analysed. For this purpose, a database-driven incident data analysis system was developed to aid the rapid review and categorization of PHMSA incident reports. Using an automated data-mining process followed by a peer review of the incident records and supported by natural hazard databases and external information sources, the pipeline Natechs were identified. As a by-product of the data-collection process, the database now includes over 800,000 incidents from all causes in industrial and transportation activities, which are automatically classified in the same way as the PHMSA record. This presentation describes the data collection and reviewing steps conducted during the study, provides information on the developed database and data analysis tools, and reports the findings of a statistical analysis of the identified hazardous liquid pipeline incidents in terms of accident dynamics and

  18. Accident Sequence Evaluation Program: Human reliability analysis procedure

    SciTech Connect

    Swain, A.D.

    1987-02-01

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

  19. Probabilistic methods for accident-progression analysis

    SciTech Connect

    Jamali, K. M.

    1981-01-01

    Probabilistic methods that can be used as basis for deterministic calculations of transients or accidents in nuclear power plants are described. They include obtaining initiator-dependent sequences on the component level and related analyses, propagation of primary event uncertainties in the ranking of sequences, and detailed treatment of dependent failures. The results are shown for protected transients in the short term forced circulation phase of decay heat removal in the Clinch River Breeder Reactor. Higher values of unavailabilities are obtained than previous works as a result of more detailed common cause/mode failure modeling. The unavailability of decay heat removal by forced circulation for the loss of off-site power and loss of main feedwater system initiators is estimated at 4 x 10/sup -3//yr and 9 x 10/sup -3//yr, respectively. 15 refs., 1 fig., 2 tabs.

  20. Accident analysis of the windowless target system

    SciTech Connect

    Bianchi, F.; Ferri, R.

    2006-07-01

    Transmutation systems are able to reduce the radio-toxicity and amount of High-Level Wastes (HLW), which are the main concerns related to the peaceful use of nuclear energy, and therefore they should make nuclear energy more easily acceptable by population. A transmutation system consists of a sub-critical fast reactor, an accelerator and a Target System, where the spallation reactions needed to sustain the chain reaction take place. Three options were proposed for the Target System within the European project PDS-XADS (Preliminary Design Studies on an Experimental Accelerator Driven System): window, windowless and solid. This paper describes the constraints taken into account in the design of the windowless Target System for the large Lead-Bismuth-Eutectic cooled XADS and deals with the results of the calculations performed to assess the behaviour of the target during some accident sequences related to pump trips. (authors)

  1. MELCOR accident analysis for ARIES-ACT

    SciTech Connect

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

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

  2. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect

    Baek, J. S.; Cheng, L. Y.; Diamond, D.

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  3. FSAR fire accident analysis for a plutonium facility

    SciTech Connect

    Lam, K.

    1997-06-01

    The Final Safety Analysis Report (FSAR) for a plutonium facility as required by DOE Orders 5480.23 and 5480.22 has recently been completed and approved. The facility processes and stores radionuclides such as Pu-238, Pu-239, enriched uranium, and to a lesser degree other actinides. This facility produces heat sources. DOE Order 5480.23 and DOE-STD-3009-94 require analysis of different types of accidents (operational accidents such as fires, explosions, spills, criticality events, and natural phenomena such as earthquakes). The accidents that were analyzed quantitatively, or the Evaluation Basis Accidents (EBAs), were selected based on a multi-step screening process that utilizes extensively the Hazards Analysis (HA) performed for the facility. In the HA, specific accident scenarios, with estimated frequency and consequences, were developed for each identified hazard associated with facility operations and activities. Analysis of the EBAs and comparison of their consequences to the evaluation guidelines established the safety envelope for the facility and identified the safety-class structures, systems, and components. This paper discusses the analysis of the fire EBA. This fire accident was analyzed in relatively great detail in the FSAR because of its potential off-site consequences are more severe compared to other events. In the following, a description of the scenario is first given, followed by a brief summary of the methodology for calculating the source term. Finally, the author discuss how a key parameter affecting the source term, the leakpath factor, was determined, which is the focus of this paper.

  4. 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.

  5. Code System for Toxic Gas Accident Analysis.

    Energy Science and Technology Software Center (ESTSC)

    2001-09-24

    Version 00 TOXRISK is an interactive program developed to aid in the evaluation of nuclear power plant control room habitability in the event of a nearby toxic material release. The program uses a model which is consistent with the approach described in the NRC Regulatory Guide 1.78. Release of the gas is treated as an initial puff followed by a continuous plume. The relative proportions of these as well as the plume release rate aremore » supplied by the user. Transport of the gas is modeled as a Gaussian distribution and occurs through the action of a constant velocity, constant direction wind. Dispersion or diffusion of the gas during transport is described by modified Pasquill-Gifford dispersion coefficients. Great flexibility is afforded the user in specifying the release description, meteorological conditions, relative geometry of the accident and plant, and the plant ventilation system characteristics. Two types of simulation can be performed: multiple case (parametric) studies and probabilistic analyses.« less

  6. Criticality accident dosimetry by chromosomal analysis.

    PubMed

    Voisin, P; Roy, L; Hone, P A; Edwards, A A; Lloyd, D C; Stephan, G; Romm, H; Groer, P G; Brame, R

    2004-01-01

    The technique of measuring the frequency of dicentric chromosomal aberrations in blood lymphocytes was used to estimate doses in a simulated criticality accident. The simulation consisted of three exposures; approximately 5 Gy with a bare source and 1 and 2 Gy with a lead-shielded source. Three laboratories made separate estimates of the doses. These were made by the iterative method of apportioning the observed dicentric frequencies between the gamma and neutron components, taking account of a given gamma/neutron dose ratio, and referring the separated dicentric frequencies to dose-response calibration curves. An alternative method, based on Bayesian ideas, was employed. This was developed for interpreting dicentric frequencies in situations where the gamma/neutron ratio is uncertain. Both methods gave very similar results. One laboratory produced dose estimates close to the eventual exercise reference doses and the other laboratories estimated slightly higher values. The main reason for the higher values was the calibration relationships for fission neutrons. PMID:15353688

  7. Severe Accident Analysis Code SAMPSON Improvement for IMPACT Project

    NASA Astrophysics Data System (ADS)

    Ujita, Hiroshi; Ikeda, Takashi; Naitoh, Masanori

    SAMPSON is the integral code for severe accident analysis in detail with modular structure, developed in the IMPACT project. Each module can run independently and communication with multiple analysis modules supervised by the analysis control module makes an integral analysis possible. At the end of Phase 1 (1994-1997), demonstration simulations by combinations of up to 11 analysis modules had been performed and physical models in the code had been verified by separate-effect tests and validated by inegral tests. Multi-dimensional mechanistic models and theoretical-based conservation equations have been applied, during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. Verificaton and validation have been performed by analysing separate-effect tests and inegral tests, while actual plant analyses are also being in progress.

  8. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  9. a Simplified Methodology for the Prediction of the Small Break Loss-Of Accident.

    NASA Astrophysics Data System (ADS)

    Ward, Leonard William

    1988-12-01

    This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can be greatly simplified leading to a very fast executing transient system blowdown code. Because of the fast execution times, the CULSETS code, or Columbia University Loss-of-Coolant Accident and System Excursion Transient Simulator code, is ideal for performing parametric studies of Emergency Core Cooling system or assessing the consequences of the many operator actions performed to place the system in a long term cooling mode following a small break LOCA. While the methodology was designed with specific application to the small break loss-of-coolant accident, it can also be used to simulate loss-of-feedwater, steam line breaks, and steam generator tube rupture events. The code is easily adaptable to a personal computer and could also be modified to provide the primary and secondary system responses to supply the required inputs to a simulator for a pressurized water reactor.

  10. INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

    SciTech Connect

    D.A. Kalinich

    1999-09-27

    Impacts due to nearby installations and operations were determined in the Preliminary MGDS Hazards Analysis (CRWMS M&O 1996) to be potentially applicable to the proposed repository at Yucca Mountain. This determination was conservatively based on limited knowledge of the potential activities ongoing on or off the Nevada Test Site (NTS). It is intended that the Industrial/Military Activity-Initiated Accident Screening Analysis provided herein will meet the requirements of the ''Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants'' (NRC 1987) in establishing whether this external event can be screened from further consideration or must be included as a design basis event (DBE) in the development of accident scenarios for the Monitored Geologic Repository (MGR). This analysis only considers issues related to preclosure radiological safety. Issues important to waste isolation as related to impact from nearby installations will be covered in the MGR performance assessment.

  11. Cold Vacuum Drying facility design basis accident analysis documentation

    SciTech Connect

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

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

    SciTech Connect

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

    2008-10-01

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

  13. 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.

  14. Analysis of Crew Fatigue in AIA Guantanamo Bay Aviation Accident

    NASA Technical Reports Server (NTRS)

    Rosekind, Mark R.; Gregory, Kevin B.; Miller, Donna L.; Co, Elizabeth L.; Lebacqz, J. Victor; Statler, Irving C. (Technical Monitor)

    1994-01-01

    Flight operations can engender fatigue, which can affect flight crew performance, vigilance, and mood. The National Transportation Safety Board (NTSB) requested the NASA Fatigue Countermeasures Program to analyze crew fatigue factors in an aviation accident that occurred at Guantanamo Bay, Cuba. There are specific fatigue factors that can be considered in such investigations: cumulative sleep loss, continuous hours of wakefulness prior to the incident or accident, and the time of day at which the accident occurred. Data from the NTSB Human Performance Investigator's Factual Report, the Operations Group Chairman's Factual Report, and the Flight 808 Crew Statements were analyzed, using conservative estimates and averages to reconcile discrepancies among the sources. Analysis of these data determined the following: the entire crew displayed cumulative sleep loss, operated during an extended period of continuous wakefulness, and obtained sleep at times in opposition to the circadian disposition for sleep, and that the accident occurred in the afternoon window of physiological sleepiness. In addition to these findings, evidence that fatigue affected performance was suggested by the cockpit voice recorder (CVR) transcript as well as in the captain's testimony. Examples from the CVR showed degraded decision-making skills, fixation, and slowed responses, all of which can be affected by fatigue; also, the captain testified to feeling "lethargic and indifferent" just prior to the accident. Therefore, the sleep/wake history data supports the hypothesis that fatigue was a factor that affected crewmembers' performance. Furthermore, the examples from the CVR and the captain's testimony support the hypothesis that the fatigue had an impact on specific actions involved in the occurrence of the accident.

  15. Code System for Best-Estimate Analysis of LOCA in BWR.

    Energy Science and Technology Software Center (ESTSC)

    2001-07-23

    Version 00 TRAC-BD1 performs best estimate analyses of loss-of-coolant accidents (LOCA) and other transients in boiling water reactors (BWRs). The program provides LOCA analysis capability for BWRs and for many BWR-related thermal-hydraulic experimental facilities. The program features a three-dimensional treatment of the BWR pressure vessel, a detailed model of a BWR fuel bundle including multi-rod, multi-bundle, radiation heat transfer, and leakage path modeling capability; flow-regime-dependent constitutive equation treatment; reflood tracking capability both for falling filmsmore » and bottom flood quench fronts; and consistent treatment of the entire accident sequence. Dump/restart capabilities are also provided.« less

  16. GASFLOW analysis of a tritium leak accident

    SciTech Connect

    Farman, R.F.; Fujita, R.K.; Travis, J.R.

    1994-09-01

    The consequences of an earthquake-induced fire involving a tritium leak were analyzed using the GASFLOW computer code. Modeling features required by the analysis include ventilation boundary conditions, flow of a gas mixture in an enclosure containing obstacles, thermally induced buoyancy, and combustion phenomena.

  17. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

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

  18. MELCOR analyses of severe accident scenarios in Oconee, a B&W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-03-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  19. MELCOR analyses of severe accident scenarios in Oconee, a B W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S. ); Foulds, R. )

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  20. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

    power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

  1. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

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

    2015-04-16

    power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

  2. 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.

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

    SciTech Connect

    Hall, B.W.

    1996-09-25

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Subsurface Leaks Resulting in Pool.

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

    SciTech Connect

    Hall, B.W.

    1996-09-25

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Surface Leaks Resulting in Pool.

  5. Small Break LOCA Analysis of ACR-700 NPP

    SciTech Connect

    Limin Zheng; Sen Shen; Wright, David

    2006-07-01

    A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied. (authors)

  6. Comparative analysis of pressure vessel integrity for various LOCA conditions

    NASA Astrophysics Data System (ADS)

    Çolak, Üner; Özdere, Oya

    2001-09-01

    In this study, integrity analysis is performed for a classical four loop PWR pressure vessel fabricated from SA533B type ferritic steel. Pressure vessel behavior is analyzed by deterministic and probabilistic methods under transient conditions, which may cause pressurized thermal shock (PTS). In deterministic analysis, the change of material properties and the mechanical state of the vessel are analyzed against changes in coolant pressure and temperature. Probabilistic analysis is performed to obtain pressure vessel beltline region weld failure probabilities in transient conditions. Overall vessel failure probabilities are evaluated based on the results of deterministic analyses. Computer code VISA-II is utilized for the calculation of vessel failure probabilities. Among three cases considered in this study, a medium break loss of coolant accident induced by a 50 cm2 break in the hot leg yields the highest vessel rupture probability. The maximum nil ductility temperature in all cases is still below the NRC PTS limit.

  7. NASA Accident Precursor Analysis Handbook, Version 1.0

    NASA Technical Reports Server (NTRS)

    Groen, Frank; Everett, Chris; Hall, Anthony; Insley, Scott

    2011-01-01

    Catastrophic accidents are usually preceded by precursory events that, although observable, are not recognized as harbingers of a tragedy until after the fact. In the nuclear industry, the Three Mile Island accident was preceded by at least two events portending the potential for severe consequences from an underappreciated causal mechanism. Anomalies whose failure mechanisms were integral to the losses of Space Transportation Systems (STS) Challenger and Columbia had been occurring within the STS fleet prior to those accidents. Both the Rogers Commission Report and the Columbia Accident Investigation Board report found that processes in place at the time did not respond to the prior anomalies in a way that shed light on their true risk implications. This includes the concern that, in the words of the NASA Aerospace Safety Advisory Panel (ASAP), "no process addresses the need to update a hazard analysis when anomalies occur" At a broader level, the ASAP noted in 2007 that NASA "could better gauge the likelihood of losses by developing leading indicators, rather than continue to depend on lagging indicators". These observations suggest a need to revalidate prior assumptions and conclusions of existing safety (and reliability) analyses, as well as to consider the potential for previously unrecognized accident scenarios, when unexpected or otherwise undesired behaviors of the system are observed. This need is also discussed in NASA's system safety handbook, which advocates a view of safety assurance as driving a program to take steps that are necessary to establish and maintain a valid and credible argument for the safety of its missions. It is the premise of this handbook that making cases for safety more experience-based allows NASA to be better informed about the safety performance of its systems, and will ultimately help it to manage safety in a more effective manner. The APA process described in this handbook provides a systematic means of analyzing candidate

  8. Civil helicopter wire strike assessment study. Volume 2: Accident analysis briefs

    NASA Technical Reports Server (NTRS)

    Tuomela, C. H.; Brennan, M. F.

    1980-01-01

    A description and analysis of each of the 208 civil helicopter wire strike accidents reported to the National Transportation Safety Board (NTSB) for the ten year period 1970-1979 is given. The accident analysis briefs were based on pilot reports, FAA investigation reports, and such accident photographs as were made available. Briefs were grouped by year and, within year, by NTSB accident report number.

  9. Road Traffic Accident Analysis of Ajmer City Using Remote Sensing and GIS Technology

    NASA Astrophysics Data System (ADS)

    Bhalla, P.; Tripathi, S.; Palria, S.

    2014-12-01

    With advancement in technology, new and sophisticated models of vehicle are available and their numbers are increasing day by day. A traffic accident has multi-facet characteristics associated with it. In India 93% of crashes occur due to Human induced factor (wholly or partly). For proper traffic accident analysis use of GIS technology has become an inevitable tool. The traditional accident database is a summary spreadsheet format using codes and mileposts to denote location, type and severity of accidents. Geo-referenced accident database is location-referenced. It incorporates a GIS graphical interface with the accident information to allow for query searches on various accident attributes. Ajmer city, headquarter of Ajmer district, Rajasthan has been selected as the study area. According to Police records, 1531 accidents occur during 2009-2013. Maximum accident occurs in 2009 and the maximum death in 2013. Cars, jeeps, auto, pickup and tempo are mostly responsible for accidents and that the occurrence of accidents is mostly concentrated between 4PM to 10PM. GIS has proved to be a good tool for analyzing multifaceted nature of accidents. While road safety is a critical issue, yet it is handled in an adhoc manner. This Study is a demonstration of application of GIS for developing an efficient database on road accidents taking Ajmer City as a study. If such type of database is developed for other cities, a proper analysis of accidents can be undertaken and suitable management strategies for traffic regulation can be successfully proposed.

  10. 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.

  11. Comprehensive Analysis of Two Downburst-Related Aircraft Accidents

    NASA Technical Reports Server (NTRS)

    Shen, J.; Parks, E. K.; Bach, R. E.

    1996-01-01

    Although downbursts have been identified as the major cause of a number of aircraft takeoff and landing accidents, only the 1985 Dallas/Fort Worth (DFW) and the more recent (July 1994) Charlotte, North Carolina, landing accidents provided sufficient onboard recorded data to perform a comprehensive analysis of the downburst phenomenon. The first step in the present analysis was the determination of the downburst wind components. Once the wind components and their gradients were determined, the degrading effect of the wind environment on the airplane's performance was calculated. This wind-shear-induced aircraft performance degradation, sometimes called the F-factor, was broken down into two components F(sub 1) and F(sub 2), representing the effect of the horizontal wind gradient and the vertical wind velocity, respectively. In both the DFW and Charlotte cases, F(sub 1) was found to be the dominant causal factor of the accident. Next, the aircraft in the two cases were mathematically modeled using the longitudinal equations of motion and the appropriate aerodynamic parameters. Based on the aircraft model and the determined winds, the aircraft response to the recorded pilot inputs showed good agreement with the onboard recordings. Finally, various landing abort strategies were studied. It was concluded that the most acceptable landing abort strategy from both an analytical and pilot's standpoint was to hold constant nose-up pitch attitude while operating at maximum engine thrust.

  12. Analysis of Three Mile Island-Unit 2 accident

    SciTech Connect

    Not Available

    1980-03-01

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

  13. System code requirements for safety analysis of SBWR

    SciTech Connect

    Andersen, J.G.M.; Shiralkar, B.S.

    1994-12-31

    The simplified boiling water reactor (SBWR) being developed by General Electric Nuclear Energy is an advanced boiling water reactor relying on natural circulation during normal operation and passive safety features. The major elements of the passive safety features are the automatic depressurization of the reactor pressure vessel (RPV) following a loss-of-coolant accident (LOCA) through safety/relief valves and depressurization valves, the gravity-driven coolant system (GDCS), and the passive containment cooling system (PCCS) for residual heat removal. These passive safety systems, although based on existing technology, have generated new requirements for the computer codes used in safety and design analysis. TRACG is the computer code used for safety and design analysis for the SBWR.

  14. 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

  15. Traffic accident analysis using GIS: a case study of Kyrenia City

    NASA Astrophysics Data System (ADS)

    Kara, Can; Akçit, Nuhcan

    2015-06-01

    Traffic accidents are causing major deaths in urban environments, so analyzing locations of the traffic accidents and their reasons is crucial. In this manner, patterns of accidents and hotspot distribution are analyzed by using geographic information technology. Locations of the traffic accidents in the years 2011, 2012 and 2013 are combined to generate the kernel distribution map of Kyrenia City. This analysis aims to find high dense intersections and segments within the city. Additionally, spatial autocorrelation methods Local Morans I and Getis-Ord Gi are employed . The results are discussed in detail for further analysis. Finally, required changes for numerous intersections are suggested to decrease potential risks of high dense accident locations.

  16. Predicting System Accidents with Model Analysis During Hybrid Simulation

    NASA Technical Reports Server (NTRS)

    Malin, Jane T.; Fleming, Land D.; Throop, David R.

    2002-01-01

    Standard discrete event simulation is commonly used to identify system bottlenecks and starving and blocking conditions in resources and services. The CONFIG hybrid discrete/continuous simulation tool can simulate such conditions in combination with inputs external to the simulation. This provides a means for evaluating the vulnerability to system accidents of a system's design, operating procedures, and control software. System accidents are brought about by complex unexpected interactions among multiple system failures , faulty or misleading sensor data, and inappropriate responses of human operators or software. The flows of resource and product materials play a central role in the hazardous situations that may arise in fluid transport and processing systems. We describe the capabilities of CONFIG for simulation-time linear circuit analysis of fluid flows in the context of model-based hazard analysis. We focus on how CONFIG simulates the static stresses in systems of flow. Unlike other flow-related properties, static stresses (or static potentials) cannot be represented by a set of state equations. The distribution of static stresses is dependent on the specific history of operations performed on a system. We discuss the use of this type of information in hazard analysis of system designs.

  17. Extension of ship accident analysis to multiple-package shipments

    SciTech Connect

    Mills, G.S.; Neuhauser, K.S.

    1997-11-01

    Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously. Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10`s or 100`s of individual packagings is compromised. The previous analysis involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship`s bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence.

  18. 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.

  19. 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)

  20. PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

    SciTech Connect

    , K; Jonathan Lowrie, J; David Thoman , D; Austin Keller , A

    2008-07-30

    Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases.

  1. Offsite radiological consequence analysis for the bounding aircraft crash accident

    SciTech Connect

    OBERG, B.D.

    2003-03-22

    The purpose of this calculation note is to quantitatively analyze a bounding aircraft crash accident for comparison to the DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', Appendix A, Evaluation Guideline of 25 rem. The potential of aircraft impacting a facility was evaluated using the approach given in DOE-STD-3014-96, ''Accident Analysis for Aircraft Crash into Hazardous Facilities''. The following aircraft crash frequencies were determined for the Tank Farms in RPP-11736, ''Assessment Of Aircraft Crash Frequency For The Hanford Site 200 Area Tank Farms'': (1) The total aircraft crash frequency is ''extremely unlikely.'' (2) The general aviation crash frequency is ''extremely unlikely.'' (3) The helicopter crash frequency is ''beyond extremely unlikely.'' (4) For the Hanford Site 200 Areas, other aircraft type, commercial or military, each above ground facility, and any other type of underground facility is ''beyond extremely unlikely.'' As the potential of aircraft crash into the 200 Area tank farms is more frequent than ''beyond extremely unlikely,'' consequence analysis of the aircraft crash is required.

  2. TRAC analysis of design basis events for the accelerator production of tritium target/blanket

    SciTech Connect

    Lin, J.C.; Elson, J.

    1997-08-01

    A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers.

  3. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    NASA Astrophysics Data System (ADS)

    Gussev, M. N.; Byun, T. S.; Yamamoto, Y.; Maloy, S. A.; Terrani, K. A.

    2015-11-01

    One of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filtering unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.

  4. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    DOE PAGESBeta

    Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Gussev, M. N.; Terrani, K. A.

    2015-08-25

    Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

  5. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    SciTech Connect

    Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Gussev, M. N.; Terrani, K. A.

    2015-08-25

    Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filtering unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.

  6. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    SciTech Connect

    Gussev, Maxim N.; Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Terrani, Kurt A.

    2015-11-01

    The high resistance of cladding to plastic deformation and burst failure is one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) since the deformation and burst behavior governs the cooling efficiency of flow channels and process of fission product release. To simulate and evaluate such deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisted of a high-resolution video camera, light filtering unit, and monochromatic light sources, and the in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. In this study eleven (11) candidate cladding materials for ATF, i.e., 6 FeCrAl alloys and 5 nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800°C while negligible strain rates were measured for higher strength alloys and/or for relatively thick wall specimens.

  7. Statistical analysis of sudden chemical leakage accidents reported in China between 2006 and 2011.

    PubMed

    Li, Yang; Ping, Hua; Ma, Zhi-Hong; Pan, Li-Gang

    2014-04-01

    According to the data from authoritative sources, 1,400 sudden leakage accidents occurred in China during 2006 to 2011 were investigated, in which, 666 accidents were used for statistical characteristic abstracted with no or little damage. The research results were as follows: (1) Time fluctuation: the yearly number of sudden leakage accidents is shown to be decreasing from 2006 to 2010, and a slightly increase in 2011. Sudden leakage accidents occur mainly in summer, and more than half of the accidents occur from May to September. (2) Regional distribution: the accidents are highly concentrated in the coastal area, in which accidents result from small and medium-sized enterprises more easily than that of the larger ones. (3) Pollutants: hazardous chemicals are up to 95 % of sudden leakage accidents. (4) Steps: transportation represents almost half of the accidents, followed by production, usage, storage, and discard. (5) Pollution and casualties: it is easy to cause environmental pollution and casualties. (6) Causes: more than half of the cases were caused by human factor, followed by management reason, and equipment failure. However, sudden chemical leakage may also be caused by high temperature, rain, wet road, and terrain. (7) The results of principal component analysis: five factors are extracted by the principal component analysis, including pollution, casualties, regional distribution, steps, and month. According to the analysis of the accident, the characteristics, causes, and damages of the sudden leakage accident will be investigated. Therefore, advices for prevention and rescue should be acquired. PMID:24407779

  8. Analysis of surface powered haulage accidents, January 1990--July 1996

    SciTech Connect

    Fesak, G.M.; Breland, R.M.; Spadaro, J.

    1996-12-31

    This report addresses surface haulage accidents that occurred between January 1990 and July 1996 involving haulage trucks (including over-the-road trucks), front-end-loaders, scrapers, utility trucks, water trucks, and other mobile haulage equipment. The study includes quarries, open pits and surface coal mines utilizing self-propelled mobile equipment to transport personnel, supplies, rock, overburden material, ore, mine waste, or coal for processing. A total of 4,397 accidents were considered. This report summarizes the major factors that led to the accidents and recommends accident prevention methods to reduce the frequency of these accidents.

  9. Decontamination analysis of the NUWAX-83 accident site using DECON

    SciTech Connect

    Tawil, J.J.

    1983-11-01

    This report presents an analysis of the site restoration options for the NUWAX-83 site, at which an exercise was conducted involving a simulated nuclear weapons accident. This analysis was performed using a computer program deveoped by Pacific Northwest Laboratory. The computer program, called DECON, was designed to assist personnel engaged in the planning of decontamination activities. The many features of DECON that are used in this report demonstrate its potential usefulness as a site restoration planning tool. Strategies that are analyzed with DECON include: (1) employing a Quick-Vac option, under which selected surfaces are vacuumed before they can be rained on; (2) protecting surfaces against precipitation; (3) prohibiting specific operations on selected surfaces; (4) requiring specific methods to be used on selected surfaces; (5) evaluating the trade-off between cleanup standards and decontamination costs; and (6) varying of the cleanup standards according to expected exposure to surface.

  10. An Accident Precursor Analysis Process Tailored for NASA Space Systems

    NASA Technical Reports Server (NTRS)

    Groen, Frank; Stamatelatos, Michael; Dezfuli, Homayoon; Maggio, Gaspare

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system and which may differ in frequency or type from those in the various models. These discrepancies between the models (perceived risk) and the system (actual risk) provide the leading indication of an underappreciated risk. This paper presents an APA process developed specifically for NASA Earth-to-Orbit space systems. The purpose of the process is to identify and characterize potential sources of system risk as evidenced by anomalous events which, although not necessarily presenting an immediate safety impact, may indicate that an unknown or insufficiently understood risk-significant condition exists in the system. Such anomalous events are considered accident precursors because they signal the potential for severe consequences that may occur in the future, due to causes that are discernible from their occurrence today. Their early identification allows them to be integrated into the overall system risk model used to intbrm decisions relating to safety.

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

    SciTech Connect

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

    1992-11-01

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

  12. Systemic analysis of so-called 'accidents on the level' in a multi trade company.

    PubMed

    Leclercq, S; Thouy, S

    2004-10-10

    Slips, trips and falls on the level are considered commonplace and are rarely subjected to in-depth analysis. They occur in highly varied circumstances in an occupational situation. In-depth analysis of these accidents was conducted within a company with the aim of understanding them better, to be able to discuss prevention field possibilities and priorities for the company concerned. Firstly, available data on 'accidents on the level' occurring over the last 4 years were analysed and a typology for these accidents was derived, based on individual activity at the time of the accident and accident location. The three most serious accident-causing situations were analysed in-depth from interviews with injured persons, as well as from activity observation and activity-related verbal information obtained from operatives. These most serious situations involved accidents occurring when climbing down from trucks or when walking either in surroundings outside company premises or from (to) a vehicle to (from) a work location. In-depth accident analysis and characterization of accident-causing situations as a whole enhance our understanding of the accident process and allow us to envisage priorities for action in the prevention field, in operational terms. Each accident-causing situation reveals environmental factors that in fact constitute accident factors (obstacle, stone, etc.), when the individual walks or climbs down from a truck. Analysis shows that other events are necessary for accident occurrence. For example, the individual may be subjected to a time constraint or may be preoccupied. Results obtained here, in a company integrating different trades, are discussed and compared with those referred to in the literature. Generalization of some of these results is also considered. PMID:15370848

  13. Uncertainty Analysis of RELAP5-3D

    SciTech Connect

    Alexandra E Gertman; Dr. George L Mesina

    2012-07-01

    As world-wide energy consumption continues to increase, so does the demand for the use of alternative energy sources, such as Nuclear Energy. Nuclear Power Plants currently supply over 370 gigawatts of electricity, and more than 60 new nuclear reactors have been commissioned by 15 different countries. The primary concern for Nuclear Power Plant operation and lisencing has been safety. The safety of the operation of Nuclear Power Plants is no simple matter- it involves the training of operators, design of the reactor, as well as equipment and design upgrades throughout the lifetime of the reactor, etc. To safely design, operate, and understand nuclear power plants, industry and government alike have relied upon the use of best-estimate simulation codes, which allow for an accurate model of any given plant to be created with well-defined margins of safety. The most widely used of these best-estimate simulation codes in the Nuclear Power industry is RELAP5-3D. Our project focused on improving the modeling capabilities of RELAP5-3D by developing uncertainty estimates for its calculations. This work involved analyzing high, medium, and low ranked phenomena from an INL PIRT on a small break Loss-Of-Coolant Accident as wall as an analysis of a large break Loss-Of- Coolant Accident. Statistical analyses were performed using correlation coefficients. To perform the studies, computer programs were written that modify a template RELAP5 input deck to produce one deck for each combination of key input parameters. Python scripting enabled the running of the generated input files with RELAP5-3D on INL’s massively parallel cluster system. Data from the studies was collected and analyzed with SAS. A summary of the results of our studies are presented.

  14. COBRA/TRAC analysis of the PKL reflood test K9. [PWR

    SciTech Connect

    Wilkins, C.A.; Thurgood, M.J.

    1982-08-01

    Experiments at the Primaerkreislaeufe (PKL) test facility in Erlangen, Germany, simulated the refill and reflood procedure after a loss-of-coolant accident (LOCA) in the primary coolant system of a 1300-MW pressurized water reactor (PWR). COBRA/TRAC, a thermal-hydraulics analysis code developed at the Pacific Northwest Laboratory, was used to model experiment K9 of the PKL test series (completed December 1979). The COBRA/TRAC code, which utilizes COBRA-TF as the vessel module and TRAC-P1A for the remaining components, was designed to analyze LOCAs in PWRs. PKL-K9 was characterized by a double-ended guillotine break in the cold leg with emergency core cooling water injected into the cold legs. COBRA/TRAC was able to successfully predict lower-core temperature profiles and quench times, upper-core temperature profiles until the quench, upper plenum and break pressures, and correct trends in collapsed water levels.

  15. Evaluation model calculations with the water reactor analysis package (WRAP-EM)

    SciTech Connect

    Gregory, M.V.; Beranek, F.

    1982-01-01

    The Water Reactor Analysis Package-Evaluation Model (WRAP-EM) is a modular system of computer codes designed to provide the safety analyst with the capability of performing complete loss-of-coolant calculations for both pressurized- and boiling-water reactor systems. The system provides a licensing-type calculation capability and thus contains most of the Nuclear Regulatory Commission-Approved EM options, as described in the Code of Federal Regulations, Title 10, Part 50, Appendix K. All phases of an accident (blowdown, refill, and reflood) are modeled. The WRAP consists of modified versions of five preexisting codes (RELAP4/MOD5, GAPCON, FRAP, MOXY, and NORCOOL), the necessary interfaces to permit automatic transition from one code to the next during the transient calculations, plus a host of user-convenience features to aid the analyst faced with a multitude of EM calculations. The WRAP has been verified against both calculated and experimental results.

  16. Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

    SciTech Connect

    Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M; Besmann, Theodore M; Hunt, Rodney Dale

    2014-01-01

    The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making the gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.

  17. An Analysis of U.S. Civil Rotorcraft Accidents by Cost and Injury (1990-1996)

    NASA Technical Reports Server (NTRS)

    Iseler, Laura; DeMaio, Joe; Rutkowski, Michael (Technical Monitor)

    2002-01-01

    A study of rotorcraft accidents was conducted to identify safety issues and research areas that might lead to a reduction in rotorcraft accidents and fatalities. The primary source of data was summaries of National Transportation Safety Board (NTSB) accident reports. From 1990 to 1996, the NTSB documented 1396 civil rotorcraft accidents in the United States in which 491 people were killed. The rotorcraft data were compared to airline and general aviation data to determine the relative safety of rotorcraft compared to other segments of the aviation industry. In depth analysis of the rotorcraft data addressed demographics, mission, and operational factors. Rotorcraft were found to have an accident rate about ten times that of commercial airliners and about the same as that of general aviation. The likelihood that an accident would be fatal was about equal for all three classes of operation. The most dramatic division in rotorcraft accidents is between flights flown by private pilots versus professional pilots. Private pilots, flying low cost aircraft in benign environments, have accidents that are due, in large part, to their own errors. Professional pilots, in contrast, are more likely to have accidents that are a result of exacting missions or use of specialized equipment. For both groups judgement error is more likely to lead to a fatal accident than are other types of causes. Several approaches to improving the rotorcraft accident rate are recommended. These mostly address improvement in the training of new pilots and improving the safety awareness of private pilots.

  18. Linguistic methodology for the analysis of aviation accidents

    NASA Technical Reports Server (NTRS)

    Goguen, J. A.; Linde, C.

    1983-01-01

    A linguistic method for the analysis of small group discourse, was developed and the use of this method on transcripts of commercial air transpot accidents is demonstrated. The method identifies the discourse types that occur and determine their linguistic structure; it identifies significant linguistic variables based upon these structures or other linguistic concepts such as speech act and topic; it tests hypotheses that support significance and reliability of these variables; and it indicates the implications of the validated hypotheses. These implications fall into three categories: (1) to train crews to use more nearly optimal communication patterns; (2) to use linguistic variables as indices for aspects of crew performance such as attention; and (3) to provide guidelines for the design of aviation procedures and equipment, especially those that involve speech.

  19. Aircraft Accident Prevention: Loss-of-Control Analysis

    NASA Technical Reports Server (NTRS)

    Kwatny, Harry G.; Dongmo, Jean-Etienne T.; Chang, Bor-Chin; Bajpai, Guarav; Yasar, Murat; Belcastro, Christine M.

    2009-01-01

    The majority of fatal aircraft accidents are associated with loss-of-control . Yet the notion of loss-of-control is not well-defined in terms suitable for rigorous control systems analysis. Loss-of-control is generally associated with flight outside of the normal flight envelope, with nonlinear influences, and with an inability of the pilot to control the aircraft. The two primary sources of nonlinearity are the intrinsic nonlinear dynamics of the aircraft and the state and control constraints within which the aircraft must operate. In this paper we examine how these nonlinearities affect the ability to control the aircraft and how they may contribute to loss-of-control. Examples are provided using NASA s Generic Transport Model.

  20. Radionuclide Analysis on Bamboos following the Fukushima Nuclear Accident

    PubMed Central

    Higaki, Takumi; Higaki, Shogo; Hirota, Masahiro; Akita, Kae; Hasezawa, Seiichiro

    2012-01-01

    In response to contamination from the recent Fukushima nuclear accident, we conducted radionuclide analysis on bamboos sampled from six sites within a 25 to 980 km radius of the Fukushima Daiichi nuclear power plant. Maximum activity concentrations of radiocesium 134Cs and 137Cs in samples from Fukushima city, 65 km away from the Fukushima Daiichi plant, were in excess of 71 and 79 kBq/kg, dry weight (DW), respectively. In Kashiwa city, 195 km away from the Fukushima Daiichi, the sample concentrations were in excess of 3.4 and 4.3 kBq/kg DW, respectively. In Toyohashi city, 440 km away from the Fukushima Daiichi, the concentrations were below the measurable limits of up to 4.5 Bq/kg DW. In the radiocesium contaminated samples, the radiocesium activity was higher in mature and fallen leaves than in young leaves, branches and culms. PMID:22496858

  1. Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

    SciTech Connect

    WILLIAMS, J.C.

    2000-09-15

    Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR).

  2. The accident analysis of mobile mine machinery in Indian opencast coal mines.

    PubMed

    Kumar, R; Ghosh, A K

    2014-01-01

    This paper presents the analysis of large mining machinery related accidents in Indian opencast coal mines. The trends of coal production, share of mining methods in production, machinery deployment in open cast mines, size and population of machinery, accidents due to machinery, types and causes of accidents have been analysed from the year 1995 to 2008. The scrutiny of accidents during this period reveals that most of the responsible factors are machine reversal, haul road design, human fault, operator's fault, machine fault, visibility and dump design. Considering the types of machines, namely, dumpers, excavators, dozers and loaders together the maximum number of fatal accidents has been caused by operator's faults and human faults jointly during the period from 1995 to 2008. The novel finding of this analysis is that large machines with state-of-the-art safety system did not reduce the fatal accidents in Indian opencast coal mines. PMID:23324038

  3. Otorhinolaryngologic disorders and diving accidents: an analysis of 306 divers.

    PubMed

    Klingmann, Christoph; Praetorius, Mark; Baumann, Ingo; Plinkert, Peter K

    2007-10-01

    Diving is a very popular leisure activity with an increasing number of participants. As more than 80% of the diving related problems involve the head and neck region, every otorhinolaryngologist should be familiar with diving medical standards. We here present an analysis of more than 300 patients we have treated in the past four years. Between January 2002 and October 2005, 306 patients presented in our department with otorhinological disorders after diving, or after diving accidents. We collected the following data: name, sex, age, date of treatment, date of accident, diagnosis, special aspects of the diagnosis, number of dives, diving certification, whether and which surgery had been performed, history of acute diving accidents or follow up treatment, assessment of fitness to dive and special remarks. The study setting was a retrospective cohort study. The distribution of the disorders was as follows: 24 divers (8%) with external ear disorders, 140 divers (46%) with middle ear disorders, 56 divers (18%) with inner ear disorders, 53 divers (17%) with disorders of the nose and sinuses, 24 divers (8%) with decompression illness (DCI) and 9 divers (3%) who complained of various symptoms. Only 18% of the divers presented with acute disorders. The most common disorder (24%) was Eustachian tube dysfunction. Female divers were significantly more often affected. Chronic sinusitis was found to be associated with a significantly higher number of performed dives. Conservative treatment failed in 30% of the patients but sinus surgery relieved symptoms in all patients of this group. The middle ear is the main problem area for divers. Middle ear ventilation problems due to Eustachian tube dysfunction can be treated conservatively with excellent results whereas pathology of the tympanic membrane and ossicular chain often require surgery. More than four out of five patients visited our department to re-establish their fitness to dive. Although the treatment of acute diving

  4. MACCS usage at Rocky Flats Plant for consequence analysis of postulated accidents

    SciTech Connect

    Foppe, T.L.; Peterson, V.L.

    1993-10-01

    The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and health effects to the public from postulated plutonium releases and from postulated criticalities. These applications were conducted to support deterministic and probabilistic accident analyses for safety analyses for safety analysis reports, radiological sabotage studies, and other regulatory requests.

  5. Assessment of RELAP5-3D for Analysis of Very High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Larry Siefken; Cliff Davis

    2005-10-01

    The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code’s capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values.

  6. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    SciTech Connect

    C. B. Davis; C. H. Oh

    2003-08-01

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide.

  7. BEACON/MOD: a computer program for thermal-hydraulic analysis of nuclear reactor containments - user's manual

    SciTech Connect

    Broadus, C.R.; Doyle, R.J.; James, S.W.; Lime, J.F.; Mings, W.J.

    1980-04-01

    The BEACON code is a best-estimate, advanced containment code designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD3, contains mass and heat transfer models for wall film and wall conduction. It is suitable for the evaluation of short-term transients in dry-containment systems. This manual describes the models employed in BEACON/MOD3 and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation.

  8. GPHS-RTG launch accident analysis for Galileo and Ulysses

    SciTech Connect

    Bradshaw, C.T. )

    1991-01-01

    This paper presents the safety program conducted to determine the response of the General Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) to potential launch accidents of the Space Shuttle for the Galileo and Ulysses missions. The National Aeronautics and Space Administration (NASA) provided definition of the Shuttle potential accidents and characterized the environments. The Launch Accident Scenario Evaluation Program (LASEP) was developed by GE to analyze the RTG response to these accidents. RTG detailed response to Solid Rocket Booster (SRB) fragment impacts, as well as to other types of impact, was obtained from an extensive series of hydrocode analyses. A comprehensive test program was conducted also to determine RTG response to the accident environments. The hydrocode response analyses coupled with the test data base provided the broad range response capability which was implemented in LASEP.

  9. An analysis of three weather-related aircraft accidents

    NASA Technical Reports Server (NTRS)

    Fujita, T. T.; Caracena, F.

    1977-01-01

    Two aircraft accidents in 1975, one at John F. Kennedy International Airport in New York City on 24 June and the other at Stapleton International Airport in Denver on 7 August, were examined in detail. A third accident on 23 June 1976 at Philadelphia International Airport is being investigated. Amazingly, there was a spearhead echo just to the north of each accident site. The echoes formed from 5 to 50 min in advance of the accident and moved faster than other echoes in the vicinity. These echoes were photographed by National Weather Service radars, 130-205 km away. At closer ranges, however, one or more circular echoes were depicted by airborne and ground radars. These cells were only 3-5 km in diameter, but they were accompanied by downdrafts of extreme intensity, called downbursts. All accidents occurred as aircraft, either descending or climbing, lost altitude while experiencing strong wind shear inside downburst cells.

  10. A POTENTIAL APPLICATION OF UNCERTAINTY ANALYSIS TO DOE-STD-3009-94 ACCIDENT ANALYSIS

    SciTech Connect

    Palmrose, D E; Yang, J M

    2007-05-10

    The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG.

  11. RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

    SciTech Connect

    Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

    1990-05-01

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

  12. Analysis of Convair 990 rejected-takeoff accident with emphasis on decision making, training and procedures

    NASA Technical Reports Server (NTRS)

    Batthauer, Byron E.

    1987-01-01

    This paper analyzes a NASA Convair 990 (CV-990) accident with emphasis on rejected-takeoff (RTO) decision making, training, procedures, and accident statistics. The NASA Aircraft Accident Investigation Board was somewhat perplexed that an aircraft could be destroyed as a result of blown tires during the takeoff roll. To provide a better understanding of tire failure RTO's, The Board obtained accident reports, Federal Aviation Administration (FAA) studies, and other pertinent information related to the elements of this accident. This material enhanced the analysis process and convinced the Accident Board that high-speed RTO's in transport aircraft should be given more emphasis during pilot training. Pilots should be made aware of various RTO situations and statistics with emphasis on failed-tire RTO's. This background information could enhance the split-second decision-making process that is required prior to initiating an RTO.

  13. Analysis of construction accidents in Turkey and responsible parties.

    PubMed

    Gürcanli, G Emre; Müngen, Uğur

    2013-01-01

    Construction is one of the world's biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972-2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00-12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

  14. Analysis of Construction Accidents in Turkey and Responsible Parties

    PubMed Central

    GÜRCANLI, G. Emre; MÜNGEN, Uğur

    2013-01-01

    Construction is one of the world’s biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972–2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00–12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

  15. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  16. A method for modeling and analysis of directed weighted accident causation network (DWACN)

    NASA Astrophysics Data System (ADS)

    Zhou, Jin; Xu, Weixiang; Guo, Xin; Ding, Jing

    2015-11-01

    Using complex network theory to analyze accidents is effective to understand the causes of accidents in complex systems. In this paper, a novel method is proposed to establish directed weighted accident causation network (DWACN) for the Rail Accident Investigation Branch (RAIB) in the UK, which is based on complex network and using event chains of accidents. DWACN is composed of 109 nodes which denote causal factors and 260 directed weighted edges which represent complex interrelationships among factors. The statistical properties of directed weighted complex network are applied to reveal the critical factors, the key event chains and the important classes in DWACN. Analysis results demonstrate that DWACN has characteristics of small-world networks with short average path length and high weighted clustering coefficient, and display the properties of scale-free networks captured by that the cumulative degree distribution follows an exponential function. This modeling and analysis method can assist us to discover the latent rules of accidents and feature of faults propagation to reduce accidents. This paper is further development on the research of accident analysis methods using complex network.

  17. Accident investigation: Analysis of aircraft motions from ATC radar recordings

    NASA Technical Reports Server (NTRS)

    Wingrove, R. C.

    1976-01-01

    A technique was developed for deriving time histories of an aircraft's motion from air traffic control (ATC) radar records. This technique uses the radar range and azimuth data, along with the downlinked altitude data (from an onboard Mode-C transponder), to derive an expanded set of data which includes airspeed, lift, thrust-drag, attitude angles (pitch, roll, and heading), etc. This method of analyzing aircraft motions was evaluated through flight experiments which used the CV-990 research aircraft and recordings from both the enroute and terminal ATC radar systems. The results indicate that the values derived from the ATC radar records are for the most part in good agreement with the corresponding values obtained from airborne measurements. In an actual accident, this analysis of ATC radar records can complement the flight-data recorders, now onboard airliners, and provide a source of recorded information for other types of aircraft that are equipped with Mode-C transponders but not with onboard recorders.

  18. Offsite Radiological Consequence Analysis for the Bounding Flammable Gas Accident

    SciTech Connect

    CARRO, C.A.

    2003-07-30

    This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a single-shell tank The calculation applies reasonably conservation input parameters in accordance with DOE-STD-3009, Appendix A, guidance. Revision 1 incorporates comments received from Office of River Protection.

  19. Accidents at work and costs analysis: a field study in a large Italian company.

    PubMed

    Battaglia, Massimo; Frey, Marco; Passetti, Emilio

    2014-01-01

    Accidents at work are still a heavy burden in social and economic terms, and action to improve health and safety standards at work offers great potential gains not only to employers, but also to individuals and society as a whole. However, companies often are not interested to measure the costs of accidents even if cost information may facilitate preventive occupational health and safety management initiatives. The field study, carried out in a large Italian company, illustrates technical and organisational aspects associated with the implementation of an accident costs analysis tool. The results indicate that the implementation (and the use) of the tool requires a considerable commitment by the company, that accident costs analysis should serve to reinforce the importance of health and safety prevention and that the economic dimension of accidents is substantial. The study also suggests practical ways to facilitate the implementation and the moral acceptance of the accounting technology. PMID:24869894

  20. Accidents at Work and Costs Analysis: A Field Study in a Large Italian Company

    PubMed Central

    BATTAGLIA, Massimo; FREY, Marco; PASSETTI, Emilio

    2014-01-01

    Accidents at work are still a heavy burden in social and economic terms, and action to improve health and safety standards at work offers great potential gains not only to employers, but also to individuals and society as a whole. However, companies often are not interested to measure the costs of accidents even if cost information may facilitate preventive occupational health and safety management initiatives. The field study, carried out in a large Italian company, illustrates technical and organisational aspects associated with the implementation of an accident costs analysis tool. The results indicate that the implementation (and the use) of the tool requires a considerable commitment by the company, that accident costs analysis should serve to reinforce the importance of health and safety prevention and that the economic dimension of accidents is substantial. The study also suggests practical ways to facilitate the implementation and the moral acceptance of the accounting technology. PMID:24869894

  1. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    SciTech Connect

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  2. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    SciTech Connect

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  3. Risk analysis of emergent water pollution accidents based on a Bayesian Network.

    PubMed

    Tang, Caihong; Yi, Yujun; Yang, Zhifeng; Sun, Jie

    2016-01-01

    To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents. PMID:26433361

  4. Waste management facility accident analysis (WASTE ACC) system: software for analysis of waste management alternatives

    SciTech Connect

    Kohout, E.F.; Folga, S.; Mueller, C.; Nabelssi, B.

    1996-03-01

    This paper describes the Waste Management Facility Accident Analysis (WASTE{underscore}ACC) software, which was developed at Argonne National Laboratory (ANL) to support the US Department of Energy`s (DOE`s) Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). WASTE{underscore}ACC is a decision support and database system that is compatible with Microsoft{reg_sign} Windows{trademark}. It assesses potential atmospheric releases from accidents at waste management facilities. The software provides the user with an easy-to-use tool to determine the risk-dominant accident sequences for the many possible combinations of process technologies, waste and facility types, and alternative cases described in the WM PEIS. In addition, its structure will allow additional alternative cases and assumptions to be tested as part of the future DOE programmatic decision-making process. The WASTE{underscore}ACC system demonstrates one approach to performing a generic, systemwide evaluation of accident risks at waste management facilities. The advantages of WASTE{underscore}ACC are threefold. First, the software gets waste volume and radiological profile data that were used to perform other WM PEIS-related analyses directly from the WASTE{underscore}MGMT system. Second, the system allows for a consistent analysis across all sites and waste streams, which enables decision makers to understand more fully the trade-offs among various policy options and scenarios. Third, the system is easy to operate; even complex scenario runs are completed within minutes.

  5. Safety analysis results for cryostat ingress accidents in ITER

    SciTech Connect

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits. 6 refs., 2 figs., 1 tab.

  6. Safety analysis results for cryostat ingress accidents in ITER

    SciTech Connect

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1996-12-31

    Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  7. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    NASA Astrophysics Data System (ADS)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  8. Action Plan for updated Chapter 15 Accident Analysis in the SRS Production Reactor SAR

    SciTech Connect

    Hightower, N.T. III; Burnett, T.W.

    1989-11-15

    This report describes the Action Plan for the upgrade of the Chapter 15 Accident Analysis in the SRS Production Reactor SAR required for K-Restart. This Action Plan will be updated periodically to reflect task accomplishments and issue resolutions.

  9. School sports accidents: analysis of causes, modes, and frequencies.

    PubMed

    Kelm, J; Ahlhelm, F; Pape, D; Pitsch, W; Engel, C

    2001-01-01

    About 5% of all school children are seriously injured during physical education every year. Because of its influence on children's attitude toward sports and the economic aspects, an evaluation of causes and medical consequences is necessary. In this study, 213 school sports accidents were investigated. Besides diagnosis, the localization of injuries, as well as the duration of the sick leave were documented. Average age of injured students was 13 years. Most of the injured students blamed themselves for the accident. The most common injuries were sprains, contusions, and fractures. Main reasons for the accidents were faults in basic motion training. Playing soccer and basketball were the most frequent reasons for injuries. The upper extremity was more frequently involved than the lower extremity. Sports physicians and teachers should work out a program outlining the individual needs and capabilities of the injured students to reintegrate them into physical education. PMID:11242243

  10. Structural Analysis for the American Airlines Flight 587 Accident Investigation: Global Analysis

    NASA Technical Reports Server (NTRS)

    Young, Richard D.; Lovejoy, Andrew E.; Hilburger, Mark W.; Moore, David F.

    2005-01-01

    NASA Langley Research Center (LaRC) supported the National Transportation Safety Board (NTSB) in the American Airlines Flight 587 accident investigation due to LaRC's expertise in high-fidelity structural analysis and testing of composite structures and materials. A Global Analysis Team from LaRC reviewed the manufacturer s design and certification procedures, developed finite element models and conducted structural analyses, and participated jointly with the NTSB and Airbus in subcomponent tests conducted at Airbus in Hamburg, Germany. The Global Analysis Team identified no significant or obvious deficiencies in the Airbus certification and design methods. Analysis results from the LaRC team indicated that the most-likely failure scenario was failure initiation at the right rear main attachment fitting (lug), followed by an unstable progression of failure of all fin-to-fuselage attachments and separation of the VTP from the aircraft. Additionally, analysis results indicated that failure initiates at the final observed maximum fin loading condition in the accident, when the VTP was subjected to loads that were at minimum 1.92 times the design limit load condition for certification. For certification, the VTP is only required to support loads of 1.5 times design limit load without catastrophic failure. The maximum loading during the accident was shown to significantly exceed the certification requirement. Thus, the structure appeared to perform in a manner consistent with its design and certification, and failure is attributed to VTP loads greater than expected.

  11. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    SciTech Connect

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  12. Accident Precursor Analysis and Management: Reducing Technological Risk Through Diligence

    NASA Technical Reports Server (NTRS)

    Phimister, James R. (Editor); Bier, Vicki M. (Editor); Kunreuther, Howard C. (Editor)

    2004-01-01

    Almost every year there is at least one technological disaster that highlights the challenge of managing technological risk. On February 1, 2003, the space shuttle Columbia and her crew were lost during reentry into the atmosphere. In the summer of 2003, there was a blackout that left millions of people in the northeast United States without electricity. Forensic analyses, congressional hearings, investigations by scientific boards and panels, and journalistic and academic research have yielded a wealth of information about the events that led up to each disaster, and questions have arisen. Why were the events that led to the accident not recognized as harbingers? Why were risk-reducing steps not taken? This line of questioning is based on the assumption that signals before an accident can and should be recognized. To examine the validity of this assumption, the National Academy of Engineering (NAE) undertook the Accident Precursors Project in February 2003. The project was overseen by a committee of experts from the safety and risk-sciences communities. Rather than examining a single accident or incident, the committee decided to investigate how different organizations anticipate and assess the likelihood of accidents from accident precursors. The project culminated in a workshop held in Washington, D.C., in July 2003. This report includes the papers presented at the workshop, as well as findings and recommendations based on the workshop results and committee discussions. The papers describe precursor strategies in aviation, the chemical industry, health care, nuclear power and security operations. In addition to current practices, they also address some areas for future research.

  13. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    SciTech Connect

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  14. IAEA Activities in the Area of Safety Analysis and Accident Management

    SciTech Connect

    Lee, S.; El-Shanawany, M.

    2006-07-01

    Safety analysis is a means of demonstrating how critical safety functions, the integrity of barriers against the release of radioactive materials, and various other safety requirements are fulfilled for a broad range of operating conditions and initiating events. Accordingly, performing safety analysis for a nuclear power plant is one of the most important safety principles. Thermal-hydraulic computer codes are extensively used worldwide for safety analysis by utilities, regulatory authorities, power plant designers and vendors, nuclear fuel companies, research organizations, and technical support organizations. Safety analysis methodology and computer codes have seen a significant development over the last two decades. This fact is also reflected in the work of the International Atomic Energy Agency (IAEA) that aims at increasing the quality and international harmonization of the approaches used in safety analysis. The paper provides an overview of activities and of examples of results obtained recently or planned in the near future in the IAEA's Division of Nuclear Installation Safety in the field of safety analysis for both design basis accidents and beyond design basis accidents as well as accident management. In this paper, specific technical guidance on the safety assessments in the IAEA Safety Standards such as safety analysis methodologies, probabilistic safety assessment, and development of accident management programmes are described. Future trends and related activities in safety analysis and accident management are also introduced. (authors)

  15. 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.

  16. 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.

  17. 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.

  18. Approaches to accident analysis in recent US Department of Energy environmental impact statements

    SciTech Connect

    Mueller, C.; Folga, S.; Nabelssi, B.

    1996-12-31

    A review of accident analyses in recent US Department of Energy (DOE) Environmental Impact Statements (EISs) was conducted to evaluate the consistency among approaches and to compare these approaches with existing DOE guidance. The review considered several components of an accident analysis: the overall scope, which in turn should reflect the scope of the EIS; the spectrum of accidents considered; the methods and assumptions used to determine frequencies or frequency ranges for the accident sequences; and the assumption and technical bases for developing radiological and chemical atmospheric source terms and for calculating the consequences of airborne releases. The review also considered the range of results generated with respect to impacts on various worker and general populations. In this paper, the findings of these reviews are presented and methods recommended for improving consistency among EISs and bringing them more into line with existing DOE guidance.

  19. Injury patterns of seniors in traffic accidents: A technical and medical analysis

    PubMed Central

    Brand, Stephan; Otte, Dietmar; Mueller, Christian Walter; Petri, Maximilian; Haas, Philipp; Stuebig, Timo; Krettek, Christian; Haasper, Carl

    2012-01-01

    AIM: To investigate the actual injury situation of seniors in traffic accidents and to evaluate the different injury patterns. METHODS: Injury data, environmental circumstances and crash circumstances of accidents were collected shortly after the accident event at the scene. With these data, a technical and medical analysis was performed, including Injury Severity Score, Abbreviated Injury Scale and Maximum Abbreviated Injury Scale. The method of data collection is named the German In-Depth Accident Study and can be seen as representative. RESULTS: A total of 4430 injured seniors in traffic accidents were evaluated. The incidence of sustaining severe injuries to extremities, head and maxillofacial region was significantly higher in the group of elderly people compared to a younger age (P < 0.05). The number of accident-related injuries was higher in the group of seniors compared to other groups. CONCLUSION: Seniors are more likely to be involved in traffic injuries and to sustain serious to severe injuries compared to other groups. PMID:23173111

  20. Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery

    PubMed Central

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

    This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005–2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc. PMID:26652689

  1. Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery.

    PubMed

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

    This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005-2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc. PMID:26652689

  2. 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

  3. [Comparative analysis of the radionuclide composition in fallout after the Chernobyl and the Fukushima accidents].

    PubMed

    Kotenko, K V; Shinkarev, S M; Abramov, Iu V; Granovskaia, E O; Iatsenko, V N; Gavrilin, Iu I; Margulis, U Ia; Garetskaia, O S; Imanaka, T; Khoshi, M

    2012-01-01

    The nuclear accident occurred at Fukushima Dai-ichi Nuclear Power Plant (NPP) (March 11, 2011) similarly to the accident at the Chernobyl NPP (April 26, 1986) is related to the level 7 of the INES. It is of interest to make an analysis of the radionuclide composition of the fallout following the both accidents. The results of the spectrometric measurements were used in that comparative analysis. Two areas following the Chernobyl accident were considered: (1) the near zone of the fallout - the Belarusian part of the central spot extended up to 60 km around the Chernobyl NPS and (2) the far zone of the fallout--the "Gomel-Mogilev" spot centered 200 km to the north-northeast of the damaged reactor. In the case of Fukushima accident the near zone up to about 60 km considered. The comparative analysis has been done with respect to refractory radionuclides (95Zr, 95Nb, 141Ce, 144Ce), as well as to the intermediate and volatile radionuclides 103Ru, 106Ru, 131I, 134Cs, 137Cs, 140La, 140Ba and the results of such a comparison have been discussed. With respect to exposure to the public the most important radionuclides are 131I and 137Cs. For the both accidents the ratios of 131I/137Cs in the considered soil samples are in the similar ranges: (3-50) for the Chernobyl samples and (5-70) for the Fukushima samples. Similarly to the Chernobyl accident a clear tendency that the ratio of 131I/137Cs in the fallout decreases with the increase of the ground deposition density of 137Cs within the trace related to a radioactive cloud has been identified for the Fukushima accident. It looks like this is a universal tendency for the ratio of 131I/137Cs versus the 137Cs ground deposition density in the fallout along the trace of a radioactive cloud as a result of a heavy accident at the NPP with radionuclides releases into the environment. This tendency is important for an objective reconstruction of 131I fallout based on the results of 137Cs measurements of soil samples carried out at

  4. Application of Latin hypercube sampling to RADTRAN 4 truck accident risk sensitivity analysis

    SciTech Connect

    Mills, G.S.; Neuhauser, K.S.; Kanipe, F.L.

    1994-12-31

    The sensitivity of calculated dose estimates to various RADTRAN 4 inputs is an available output for incident-free analysis because the defining equations are linear and sensitivity to each variable can be calculated in closed mathematical form. However, the necessary linearity is not characteristic of the equations used in calculation of accident dose risk, making a similar tabulation of sensitivity for RADTRAN 4 accident analysis impossible. Therefore, a study of sensitivity of accident risk results to variation of input parameters was performed using representative routes, isotopic inventories, and packagings. It was determined that, of the approximately two dozen RADTRAN 4 input parameters pertinent to accident analysis, only a subset of five or six has significant influence on typical analyses or is subject to random uncertainties. These five or six variables were selected as candidates for Latin Hypercube Sampling applications. To make the effect of input uncertainties on calculated accident risk more explicit, distributions and limits were determined for two variables which had approximately proportional effects on calculated doses: Pasquill Category probability (PSPROB) and link population density (LPOPD). These distributions and limits were used as input parameters to Sandia`s Latin Hypercube Sampling code to generate 50 sets of RADTRAN 4 input parameters used together with point estimates of other necessary inputs to calculate 50 observations of estimated accident dose risk.Tabulations of the RADTRAN 4 accident risk input variables and their influence on output plus illustrative examples of the LHS calculations, for truck transport situations that are typical of past experience, will be presented .

  5. Synthesis of quantitative and qualitative evidence for accident analysis in risk-based highway planning.

    PubMed

    Lambert, James H; Peterson, Kenneth D; Joshi, Nilesh N

    2006-09-01

    Accident analysis involves the use of both quantitative and qualitative data in decision-making. The aim of this paper is to demonstrate the synthesis of relevant quantitative and qualitative evidence for accident analysis and for planning a large and diverse portfolio of highway investment projects. The proposed analysis and visualization techniques along with traditional mathematical modeling serve as an aid to planners, engineers, and the public in comparing the benefits of current and proposed improvement projects. The analysis uses data on crash rates, average daily traffic, cost estimates from highway agency databases, and project portfolios for regions and localities. It also utilizes up to two motivations out of seven that are outlined in the Transportation Equity Act for the 21st Century (TEA-21). Three case studies demonstrate the risk-based approach to accident analysis for short- and long-range transportation plans. The approach is adaptable to other topics in accident analysis and prevention that involve the use of quantitative and qualitative evidence, risk analysis, and multi-criteria decision-making for project portfolio selection. PMID:16730627

  6. 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.

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

    SciTech Connect

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

    2012-03-18

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

  8. Accident analysis of railway transportation of low-level radioactive and hazardous chemical wastes: Application of the /open quotes/Maximum Credible Accident/close quotes/ concept

    SciTech Connect

    Ricci, E.; McLean, R.B.

    1988-09-01

    The maximum credible accident (MCA) approach to accident analysis places an upper bound on the potential adverse effects of a proposed action by using conservative but simplifying assumptions. It is often used when data are lacking to support a more realistic scenario or when MCA calculations result in acceptable consequences. The MCA approach can also be combined with realistic scenarios to assess potential adverse effects. This report presents a guide for the preparation of transportation accident analyses based on the use of the MCA concept. Rail transportation of contaminated wastes is used as an example. The example is the analysis of the environmental impact of the potential derailment of a train transporting a large shipment of wastes. The shipment is assumed to be contaminated with polychlorinated biphenyls and low-level radioactivities of uranium and technetium. The train is assumed to plunge into a river used as a source of drinking water. The conclusions from the example accident analysis are based on the calculation of the number of foreseeable premature cancer deaths the might result as a consequence of this accident. These calculations are presented, and the reference material forming the basis for all assumptions and calculations is also provided.

  9. Analysis of fission product release behavior during the TMI-2 accident

    SciTech Connect

    Petti, D. A.; Adams, J. P.; Anderson, J. L.; Hobbins, R. R.

    1987-01-01

    An analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been initiated to provide an understanding of fission product behavior that is consistent with both the best estimate accident scenario and fission product results from the ongoing sample acquisition and examination efforts. ''First principles'' fission product release models are used to describe release from intact, disrupted, and molten fuel. Conclusions relating to fission product release, transport, and chemical form are drawn. 35 refs., 12 figs., 7 tabs.

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

    SciTech Connect

    Not Available

    1988-12-15

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

  11. Preliminary Analysis of Aircraft Loss of Control Accidents: Worst Case Precursor Combinations and Temporal Sequencing

    NASA Technical Reports Server (NTRS)

    Belcastro, Christine M.; Groff, Loren; Newman, Richard L.; Foster, John V.; Crider, Dennis H.; Klyde, David H.; Huston, A. McCall

    2014-01-01

    Aircraft loss of control (LOC) is a leading cause of fatal accidents across all transport airplane and operational classes, and can result from a wide spectrum of hazards, often occurring in combination. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of conditions and uncertainties, including multiple hazards, and their validation must provide a means of assessing system effectiveness and coverage of these hazards. This requires the definition of a comprehensive set of LOC test scenarios based on accident and incident data as well as future risks. This paper defines a comprehensive set of accidents and incidents over a recent 15 year period, and presents preliminary analysis results to identify worst-case combinations of causal and contributing factors (i.e., accident precursors) and how they sequence in time. Such analyses can provide insight in developing effective solutions for LOC, and form the basis for developing test scenarios that can be used in evaluating them. Preliminary findings based on the results of this paper indicate that system failures or malfunctions, crew actions or inactions, vehicle impairment conditions, and vehicle upsets contributed the most to accidents and fatalities, followed by inclement weather or atmospheric disturbances and poor visibility. Follow-on research will include finalizing the analysis through a team consensus process, defining future risks, and developing a comprehensive set of test scenarios with correlation to the accidents, incidents, and future risks. Since enhanced engineering simulations are required for batch and piloted evaluations under realistic LOC precursor conditions, these test scenarios can also serve as a high-level requirement for defining the engineering simulation enhancements needed for generating them.

  12. Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines

    PubMed Central

    Baka, Aikaterini D.; Uzunoglu, Nikolaos K.

    2014-01-01

    Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury) that occurred because the power supply was reconnected inadvertently or by mistake. PMID:25379331

  13. Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines.

    PubMed

    Baka, Aikaterini D; Uzunoglu, Nikolaos K

    2014-09-01

    Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury) that occurred because the power supply was reconnected inadvertently or by mistake. PMID:25379331

  14. Severe Accident Test Station Design Document

    SciTech Connect

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-09-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  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. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk.

  17. Implementation of numerical simulation techniques in analysis of the accidents in complex technological systems

    SciTech Connect

    Klishin, G.S.; Seleznev, V.E.; Aleoshin, V.V.

    1997-12-31

    Gas industry enterprises such as main pipelines, compressor gas transfer stations, gas extracting complexes belong to the energy intensive industry. Accidents there can result into the catastrophes and great social, environmental and economic losses. Annually, according to the official data several dozens of large accidents take place at the pipes in the USA and Russia. That is why prevention of the accidents, analysis of the mechanisms of their development and prediction of their possible consequences are acute and important tasks nowadays. The accidents reasons are usually of a complicated character and can be presented as a complex combination of natural, technical and human factors. Mathematical and computer simulations are safe, rather effective and comparatively inexpensive methods of the accident analysis. It makes it possible to analyze different mechanisms of a failure occurrence and development, to assess its consequences and give recommendations to prevent it. Besides investigation of the failure cases, numerical simulation techniques play an important role in the treatment of the diagnostics results of the objects and in further construction of mathematical prognostic simulations of the object behavior in the period of time between two inspections. While solving diagnostics tasks and in the analysis of the failure cases, the techniques of theoretical mechanics, of qualitative theory of different equations, of mechanics of a continuous medium, of chemical macro-kinetics and optimizing techniques are implemented in the Conversion Design Bureau {number_sign}5 (DB{number_sign}5). Both universal and special numerical techniques and software (SW) are being developed in DB{number_sign}5 for solution of such tasks. Almost all of them are calibrated on the calculations of the simulated and full-scale experiments performed at the VNIIEF and MINATOM testing sites. It is worth noting that in the long years of work there has been established a fruitful and effective

  18. Probabilistic analysis of accident precursors in the nuclear industry.

    PubMed

    Hulsmans, M; De Gelder, P

    2004-07-26

    Feedback of operating experience has always been an important issue in the nuclear industry. A probabilistic safety analysis (PSA) can be used as a tool to analyse how an operational event might have developed adversely in order to obtain a quantitative assessment of the safety significance of the event. This process is called PSA-based event analysis (PSAEA). A comprehensive set of PSAEA guidelines was developed by an international project. The main characteristics of this methodology are summarised. This approach to analyse incidents can be used to meet different objectives of utilities or nuclear regulators. The paper describes the main objectives and the experiences of the Belgian nuclear regulatory organisation AVN with the application of PSA-based event analysis. Some interesting aspects of the process of PSAEA are further developed and underlined. Several case studies are discussed and an overview of the obtained results is given. Finally, the interest of a broad and interactive forum on PSAEA is highlighted. PMID:15231351

  19. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  20. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  1. Combining task analysis and fault tree analysis for accident and incident analysis: a case study from Bulgaria.

    PubMed

    Doytchev, Doytchin E; Szwillus, Gerd

    2009-11-01

    Understanding the reasons for incident and accident occurrence is important for an organization's safety. Different methods have been developed to achieve this goal. To better understand the human behaviour in incident occurrence we propose an analysis concept that combines Fault Tree Analysis (FTA) and Task Analysis (TA). The former method identifies the root causes of an accident/incident, while the latter analyses the way people perform the tasks in their work environment and how they interact with machines or colleagues. These methods were complemented with the use of the Human Error Identification in System Tools (HEIST) methodology and the concept of Performance Shaping Factors (PSF) to deepen the insight into the error modes of an operator's behaviour. HEIST shows the external error modes that caused the human error and the factors that prompted the human to err. To show the validity of the approach, a case study at a Bulgarian Hydro power plant was carried out. An incident - the flooding of the plant's basement - was analysed by combining the afore-mentioned methods. The case study shows that Task Analysis in combination with other methods can be applied successfully to human error analysis, revealing details about erroneous actions in a realistic situation. PMID:19819365

  2. Fire Risk Analysis for Armenian NPP Confinement

    SciTech Connect

    Poghosyan, Shahen; Malkhasyan, Albert; Bznuni, Surik; Amirjanyan, Armen

    2006-07-01

    Major fire occurred at Armenian NPP (ANPP) in October 1982 showed that fire-induced initiating events (IE) can have dominant contribution in overall risk of core damage. Probabilistic Safety Assessment study for fire-induced initiating events for ANPP was initiated in 2002. Analysis was performed for compartments fires in which could result in failure of components which are necessary for reactor cold shutdown. Analysis shows that main risk from fire at ANPP is conditioned by fire in cable tunnels 61-64. Meanwhile fire in confinement compartments don't have significant contribution to overall risk of core damage. The exception is so called 'confinement valves compartment' (room no.A-013/2) fire (more than 7.5% of CDF) in which fire could result in the loss of coolant accident with unavailability of primary makeup system, which directly leads to core damage. Detailed analysis of this problem that is common for typical WWER-440/230 reactors with no hermetic MCPs and recommendations for solution are presented in this paper. (authors)

  3. 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)

  4. 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.

  5. Human and organisational factors in maritime accidents: analysis of collisions at sea using the HFACS.

    PubMed

    Chauvin, Christine; Lardjane, Salim; Morel, Gaël; Clostermann, Jean-Pierre; Langard, Benoît

    2013-10-01

    Over the last decade, the shipping industry has implemented a number of measures aimed at improving its safety level (such as new regulations or new forms of team training). Despite this evolution, shipping accidents, and particularly collisions, remain a major concern. This paper presents a modified version of the Human Factors Analysis and Classification System, which has been adapted to the maritime context and used to analyse human and organisational factors in collisions reported by the Marine Accident and Investigation Branch (UK) and the Transportation Safety Board (Canada). The analysis shows that most collisions are due to decision errors. At the precondition level, it highlights the importance of the following factors: poor visibility and misuse of instruments (environmental factors), loss of situation awareness or deficit of attention (conditions of operators), deficits in inter-ship communications or Bridge Resource Management (personnel factors). At the leadership level, the analysis reveals the frequent planning of inappropriate operations and non-compliance with the Safety Management System (SMS). The Multiple Accident Analysis provides an important finding concerning three classes of accidents. Inter-ship communications problems and Bridge Resource Management deficiencies are closely linked to collisions occurring in restricted waters and involving pilot-carrying vessels. Another class of collisions is associated with situations of poor visibility, in open sea, and shows deficiencies at every level of the socio-technical system (technical environment, condition of operators, leadership level, and organisational level). The third class is characterised by non-compliance with the SMS. This study shows the importance of Bridge Resource Management for situations of navigation with a pilot on board in restricted waters. It also points out the necessity to investigate, for situations of navigation in open sea, the masters' decisions in critical conditions

  6. 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.

  7. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion.

  8. An association between dietary habits and traffic accidents in patients with chronic liver disease: A data-mining analysis

    PubMed Central

    KAWAGUCHI, TAKUMI; SUETSUGU, TAKURO; OGATA, SHYOU; IMANAGA, MINAMI; ISHII, KUMIKO; ESAKI, NAO; SUGIMOTO, MASAKO; OTSUYAMA, JYURI; NAGAMATSU, AYU; TANIGUCHI, EITARO; ITOU, MINORU; ORIISHI, TETSUHARU; IWASAKI, SHOKO; MIURA, HIROKO; TORIMURA, TAKUJI

    2016-01-01

    The incidence of traffic accidents in patients with chronic liver disease (CLD) is high in the USA. However, the characteristics of patients, including dietary habits, differ between Japan and the USA. The present study investigated the incidence of traffic accidents in CLD patients and the clinical profiles associated with traffic accidents in Japan using a data-mining analysis. A cross-sectional study was performed and 256 subjects [148 CLD patients (CLD group) and 106 patients with other digestive diseases (disease control group)] were enrolled; 2 patients were excluded. The incidence of traffic accidents was compared between the two groups. Independent factors for traffic accidents were analyzed using logistic regression and decision-tree analyses. The incidence of traffic accidents did not differ between the CLD and disease control groups (8.8 vs. 11.3%). The results of the logistic regression analysis showed that yoghurt consumption was the only independent risk factor for traffic accidents (odds ratio, 0.37; 95% confidence interval, 0.16–0.85; P=0.0197). Similarly, the results of the decision-tree analysis showed that yoghurt consumption was the initial divergence variable. In patients who consumed yoghurt habitually, the incidence of traffic accidents was 6.6%, while that in patients who did not consume yoghurt was 16.0%. CLD was not identified as an independent factor in the logistic regression and decision-tree analyses. In conclusion, the difference in the incidence of traffic accidents in Japan between the CLD and disease control groups was insignificant. Furthermore, yoghurt consumption was an independent negative risk factor for traffic accidents in patients with digestive diseases, including CLD. PMID:27123257

  9. Space Shuttle Columbia Post-Accident Analysis and Investigation

    NASA Technical Reports Server (NTRS)

    McDanels, Steven J.

    2006-01-01

    Although the loss of the Space Shuttle Columbia and its crew was tragic, the circumstances offered a unique opportunity to examine a multitude of components which had experienced one of the harshest environments ever encountered by engineered materials: a break up at a velocity in excess of Mach 18 and an altitude exceeding 200,000 feet (63 KM), resulting in a debris field 645 miles/l,038 KM long and 10 miles/16 KM wide. Various analytical tools were employed to ascertain the sequence of events leading to the disintegration of the Orbiter and to characterize the features of the debris. The testing and analyses all indicated that a breach in a left wing reinforced carbon/carbon composite leading edge panel was the access point for hot gasses generated during re-entry to penetrate the structure of the vehicle and compromise the integrity of the materials and components in that area of the Shuttle. The analytical and elemental testing utilized such techniques as X-Ray Diffraction (XRD), Energy Dispersive X-Ray (EDX) dot mapping, Electron Micro Probe Analysis (EMPA), and X-Ray Photoelectron Spectroscopy (XPS) to characterize the deposition of intermetallics adjacent to the suspected location of the plasma breach in the leading edge of the left wing, Fig. 1.

  10. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    SciTech Connect

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, water ingestion dose, milk growing season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meat, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of 1-131 at which disposal of crops will be initiated due to accidents that occur during the growing season.

  11. Speech analysis as an index of alcohol intoxication--the Exxon Valdez accident.

    PubMed

    Brenner, M; Cash, J R

    1991-09-01

    As part of its investigation of the EXXON VALDEZ tankship accident and oil spill, the National Transportation Safety Board (NTSB) examined the master's speech for alcohol-related effects. Recorded speech samples were obtained from marine radio communications tapes. The samples were tested for four effects associated with alcohol consumption is available scientific literature: slowed speech, speech errors, misarticulation of difficult sounds ("slurring"), and audible changes in speech quality. It was found that speech immediately before and after the accident displayed large changes of the sort associated with alcohol consumption. These changes were not readily explained by fatigue, psychological stress, drug effects, or medical problems. Speech analysis appears to be a useful technique to provide secondary evidence of alcohol impairment. PMID:1930083

  12. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 1: Main report

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harrison, J.D.; Harper, F.T.; Hora, S.C.

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models.

  13. Thermodynamic analysis of cesium and iodine behavior in severe light water reactor accidents

    NASA Astrophysics Data System (ADS)

    Minato, Kazuo

    1991-11-01

    In order to understand the release and transport behavior of cesium (Cs) and iodine (I) in severe light water reactor accidents, chemical forms of Cs and I in steam-hydrogen mixtures were analyzed thermodynamically. In the calculations reactions of boron (B) with Cs were taken into consideration. The analysis showed that B plays an important role in determining chemical forms of Cs. The main Cs-containing species are CsBO 2(g) and CsBO 2(l), depending on temperature. The contribution of CsOH(g) is minor. The main I-containing species are HI(g) and CsI(g) over the wide ranges of the parameters considered. Calculations were also carried out under the conditions of the Three Mile Island Unit 2 accident.

  14. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harper, F.T.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models.

  15. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Haskin, F.E.; Harper, F.T.; Goossens, L.H.J.; Kraan, B.C.P.; Grupa, J.B.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models.

  16. Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

    SciTech Connect

    Vierow, Karen

    2005-08-29

    This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

  17. Safety culture and accident analysis--a socio-management approach based on organizational safety social capital.

    PubMed

    Rao, Suman

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, the key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization--seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly. PMID:16911855

  18. Analysis of pedestrian accident costs in Sudan using the willingness-to-pay method.

    PubMed

    Mofadal, Adam I A; Kanitpong, Kunnawee; Jiwattanakulpaisarn, Piyapong

    2015-05-01

    The willingness-to-pay (WTP) with contingent valuation (CV) method has been proven to be a valid tool for the valuation of non-market goods or socio-economic costs of road traffic accidents among communities in developed and developing countries. Research on accident costing tends to estimate the value of statistical life (VOSL) for all road users by providing a principle for the evaluation of road safety interventions in cost-benefit analysis. As in many other developing countries, the economic loss of traffic accidents in Sudan is noticeable; however, analytical research to estimate the magnitude and impact of that loss is lacking. Reports have shown that pedestrians account for more than 40% of the total number of fatalities. In this study, the WTP-CV approach was used to determine the amount of money that pedestrians in Sudan are willing to pay to reduce the risk of their own death. The impact of the socioeconomic factors, risk levels, and walking behaviors of pedestrians on their WTP for fatality risk reduction was also evaluated. Data were collected from two cities-Khartoum and Nyala-using a survey questionnaire that included 1400 respondents. The WTP-CV Payment Card Questionnaire was designed to ensure that Sudan pedestrians can easily determine the amount of money that would be required to reduce the fatality risk from a pedestrian-related accident. The analysis results show that the estimated VOSL for Sudanese pedestrians ranges from US$0.019 to US$0.101 million. In addition, the willingness-to-pay by Sudanese pedestrians to reduce their fatality risk tends to increase with age, household income, educational level, safety perception, and average time spent on social activities with family and community. PMID:25794921

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

    SciTech Connect

    Hall, B.W.

    1996-09-25

    This document contains the detailed calculations that support the spray leak accident analysis in the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR). The consequence analyses in this document form the basis for the selection of controls to mitigate or prevent spray leaks throughout TWRS. Pressurized spray leaks can occur due to a breach in containment barriers along transfer routes, during waste transfers. Spray leaks are of particular safety concern because, depending on leak dimensions, and waste pressure, they can be relatively efficient generators of dispersible sized aerosols that can transport downwind to onsite and offsite receptors. Waste is transferred between storage tanks and between processing facilities and storage tanks in TWRS through a system of buried transfer lines. Pumps for transferring waste and jumpers and valves for rerouting waste are located inside below grade pits and structures that are normally covered. Pressurized spray leaks can emanate to the atmosphere due to breaches in waste transfer associated equipment inside these structures should the structures be uncovered at the time of the leak. Pressurized spray leaks can develop through holes or cracks in transfer piping, valve bodies or pump casings caused by such mechanisms as corrosion, erosion, thermal stress, or water hammer. Leaks through degraded valve packing, jumper gaskets, or pump seals can also result in pressurized spray releases. Mechanisms that can degrade seals, packing and gaskets include aging, radiation hardening, thermal stress, etc. An1782other common cause for spray leaks inside transfer enclosures are misaligned jumpers caused by human error. A spray leak inside a DST valve pit during a transfer of aging waste was selected as the bounding, representative accident for detailed analysis. Sections 2 through 5 below develop this representative accident using the DOE- STD-3009 format. Sections 2 describes the unmitigated and mitigated accident

  20. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    PubMed

    Zhang, Yingyu; Shao, Wei; Zhang, Mengjia; Li, Hejun; Yin, Shijiu; Xu, Yingjun

    2016-07-01

    Mining has been historically considered as a naturally high-risk industry worldwide. Deaths caused by coal mine accidents are more than the sum of all other accidents in China. Statistics of 320 coal mine accidents in Shandong province show that all accidents contain indicators of "unsafe conditions of the rules and regulations" with a frequency of 1590, accounting for 74.3% of the total frequency of 2140. "Unsafe behaviors of the operator" is another important contributory factor, which mainly includes "operator error" and "venturing into dangerous places." A systems analysis approach was applied by using structural equation modeling (SEM) to examine the interactions between the contributory factors of coal mine accidents. The analysis of results leads to three conclusions. (i) "Unsafe conditions of the rules and regulations," affect the "unsafe behaviors of the operator," "unsafe conditions of the equipment," and "unsafe conditions of the environment." (ii) The three influencing factors of coal mine accidents (with the frequency of effect relation in descending order) are "lack of safety education and training," "rules and regulations of safety production responsibility," and "rules and regulations of supervision and inspection." (iii) The three influenced factors (with the frequency in descending order) of coal mine accidents are "venturing into dangerous places," "poor workplace environment," and "operator error." PMID:27085591

  1. 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

  2. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    SciTech Connect

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ``MELCOR Verification, Benchmarking, and Applications,`` whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR.

  3. Improved methodology for integral analysis of advanced reactors employing passive safety

    NASA Astrophysics Data System (ADS)

    Muftuoglu, A. Kursad

    After four decades of experience with pressurized water reactors, a new generation of nuclear plants are emerging. These advanced designs employ passive safety which relies on natural forces, such as gravity and natural circulation. The new concept of passive safety also necessitates improvement in computational tools available for best-estimate analyses. The system codes originally designed for high pressure conditions in the presence of strong momentum sources such as pumps are challenged in many ways. Increased interaction of the primary system with the containment necessitates a tool for integral analysis. This study addresses some of these concerns. An improved tool for integral analysis coupling primary system with containment calculation is also presented. The code package is based on RELAP5 and CONTAIN programs, best-estimate thermal-hydraulics code for primary system analysis and containment code for containment analysis, respectively. The suitability is demonstrated with a postulated small break loss of coolant accident analysis of Westinghouse AP600 plant. The thesis explains the details of the analysis including the coupling model.

  4. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    SciTech Connect

    Reyes, S; Gomez del Rio, J; Sanz, J

    2000-02-23

    Previous studies of the safety and environmental (S and E) aspects of the HYLIFE-II inertial fusion energy (IFE) power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work a set of computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) has been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here the authors consider a severe lost of coolant accident (LOCA) producing simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the containment) and of the two barriers surrounding the chamber (inner shielding and containment building it self). Even though containment failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. The results of these calculations show that the estimated off-site dose is less than 6 mSv (0.6 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  5. Light-Weight Radioisotope Heater Unit final safety analysis report (LWRHU-FSAR): Volume 2: Accident Model Document (AMD)

    SciTech Connect

    Johnson, E.W.

    1988-10-01

    The purpose of this volume of the LWRHU SAR, the Accident Model Document (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; Provide estimates of occurrence probabilities associated with these various accidents; Evaluate the response of the LWRHU (or its components) to the resultant accident environments; and Associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere.

  6. Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications

    SciTech Connect

    VINCENT, ANDREW

    2005-04-25

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (''Quality Assurance for Safety-Related Software'') identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls to prevent or mitigate potential accidents. Over the last year, DOE has begun several processes and programs as part of the Implementation Plan commitments, and in particular, has made significant progress in addressing several sets of issues particularly important in the application of software for performing hazard and accident analysis. The work discussed here demonstrates that through these actions, Software Quality Assurance (SQA) guidance and software tools are available that can be used to improve resulting safety analysis. Specifically, five of the primary actions corresponding to the commitments made in the Implementation Plan to Recommendation 2002-1 are identified and discussed in this paper. Included are the web-based DOE SQA Knowledge Portal and the Central Registry, guidance and gap analysis reports, electronic bulletin board and discussion forum, and a DOE safety software guide. These SQA products can benefit DOE safety contractors in the development of hazard and accident analysis by precluding inappropriate software applications and utilizing best practices when incorporating software results to safety basis documentation. The improvement actions discussed here mark a beginning to establishing stronger, standard-compliant programs, practices, and processes in SQA among safety software users, managers, and reviewers throughout the DOE Complex. Additional effort is needed, however, particularly in: (1) processes to add new software applications to the DOE Safety Software Toolbox; (2) improving the effectiveness of software issue communication; and (3) promoting a safety software quality assurance culture.

  7. Analysis of National Major Work Safety Accidents in China, 2003–2012

    PubMed Central

    YE, Yunfeng; ZHANG, Siheng; RAO, Jiaming; WANG, Haiqing; LI, Yang; WANG, Shengyong; DONG, Xiaomei

    2016-01-01

    Background: This study provides a national profile of major work safety accidents in China, which cause more than 10 fatalities per accident, intended to provide scientific basis for prevention measures and strategies to reduce major work safety accidents and deaths. Methods: Data from 2003–2012 Census of major work safety accidents were collected from State Administration of Work Safety System (SAWS). Published literature and statistical yearbook were also included to implement information. We analyzed the frequency of accidents and deaths, trend, geographic distribution and injury types. Additionally, we discussed the severity and urgency of emergency rescue by types of accidents. Results: A total of 877 major work safety accidents were reported, resulting in 16,795 deaths and 9,183 injuries. The numbers of accidents and deaths, mortality rate and incidence of major accidents have declined in recent years. The mortality rate and incidence was 0.71 and 1.20 per 106 populations in 2012, respectively. Transportation and mining contributed to the highest number of major accidents and deaths. Major aviation and railway accidents caused more casualties per incident, while collapse, machinery, electrical shock accidents and tailing dam accidents were the most severe situation that resulted in bigger proportion of death. Conclusion: Ten years’ major work safety accident data indicate that the frequency of accidents and number of eaths was declined and several safety concerns persist in some segments. PMID:27057515

  8. Trace Assessment for BWR ATWS Analysis

    SciTech Connect

    Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

    2010-04-22

    A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtained from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.

  9. Best-Estimate Analysis PWR LOCA.

    Energy Science and Technology Software Center (ESTSC)

    2005-11-11

    Version: 00 TRAC‑PF1 performs best estimate analyses of loss of coolant accidents and other transients in pressurized light water reactors. The program can also be used to model a wide range of thermal hydraulic experiments in reduced scale facilities. Models employed include reflood, multi‑dimensional two‑phase flow, nonequilibrium thermodynamics, generalized heat transfer, and reactor kinetics. Automatic steady‑state and dump/restart capabilities are provided. The changes reported in TRACNEWS issues through Number 7 are incorporated in this release.more » TRAC-PF1 was developed on a CDC computer at Los Alamos National Laboratory. The PC version of TRAC‑PF1 was converted at CNEN in 1989 and has not been updated since that time. The NRC no longer supports the TRAC codes. They currently develop and maintain the TRACE code system, which is the TRAC/RELAP Advanced Computational Engine. TRACE is a modernized thermal-hydraulics code designed to consolidate the capabilities of NRC's 3 legacy safety codes - TRAC-P, TRAC-B and RELAP. This is NRC's flagship thermal-hydraulics analysis tool. See the website for more information http://www.nrccodes.com/.« less

  10. Best-Estimate Analysis PWR LOCA.

    SciTech Connect

    MAHAFFY, J. H.

    2005-11-11

    Version: 00 TRAC‑PF1 performs best estimate analyses of loss of coolant accidents and other transients in pressurized light water reactors. The program can also be used to model a wide range of thermal hydraulic experiments in reduced scale facilities. Models employed include reflood, multi‑dimensional two‑phase flow, nonequilibrium thermodynamics, generalized heat transfer, and reactor kinetics. Automatic steady‑state and dump/restart capabilities are provided. The changes reported in TRACNEWS issues through Number 7 are incorporated in this release. TRAC-PF1 was developed on a CDC computer at Los Alamos National Laboratory. The PC version of TRAC‑PF1 was converted at CNEN in 1989 and has not been updated since that time. The NRC no longer supports the TRAC codes. They currently develop and maintain the TRACE code system, which is the TRAC/RELAP Advanced Computational Engine. TRACE is a modernized thermal-hydraulics code designed to consolidate the capabilities of NRC's 3 legacy safety codes - TRAC-P, TRAC-B and RELAP. This is NRC's flagship thermal-hydraulics analysis tool. See the website for more information http://www.nrccodes.com/.

  11. 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.

  12. Light-Weight Radioisotope Heater Unit Safety Analysis Report (LWRHU-SAR). Volume II. Accident model document

    SciTech Connect

    Johnson, E.W.

    1985-10-01

    Purposes of this volume (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; provide estimates of occurrence probabilities associated with these various accidents; evaluate the response of the LWRHU (or its components) to the resultant accident environments; and associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere.

  13. A Look at Aircraft Accident Analysis in the Early Days: Do Early 20th Century Accident Investigation Techniques Have Any Lessons for Today?

    NASA Technical Reports Server (NTRS)

    Holloway, C. M.; Johnson, C. W.

    2007-01-01

    In the early years of powered flight, the National Advisory Committee on Aeronautics in the United States produced three reports describing a method of analysis of aircraft accidents. The first report was published in 1928; the second, which was a revision of the first, was published in 1930; and the third, which was a revision and update of the second, was published in 1936. This paper describes the contents of these reports, and compares the method of analysis proposed therein to the methods used today.

  14. Hypothetical accident condition thermal analysis and testing of a Type B drum package

    SciTech Connect

    Hensel, S.J.; Alstine, M.N. Van; Gromada, R.J.

    1995-07-01

    A thermophysical property model developed to analytically determine the thermal response of cane fiberboard when exposed to temperatures and heat fluxes associated with the 10 CFR 71 hypothetical accident condition (HAC) has been benchmarked against two Type B drum package fire test results. The model 9973 package was fire tested after a 30 ft. top down drop and puncture, and an undamaged model 9975 package containing a heater (21W) was fire tested to determine content heat source effects. Analysis results using a refined version of a previously developed HAC fiberboard model compared well against the test data from both the 9973 and 9975 packages.

  15. Accident sequence analysis for a BWR (Boiling Water Reactor) during low power and shutdown operations

    SciTech Connect

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs.

  16. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect

    Pasichnyk, I.; Perin, Y.; Velkov, K.

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  17. Quantifying reactor safety margins: Application of CSAU (Code Scalability, Applicability and Uncertainty) methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

    SciTech Connect

    Wulff, W.; Boyack, B.E.; Duffey, R.B.; Griffith, P.; Katsma, K.R.; Lellouche, G.S.; Levy, S.; Rohatgi, U.S.; Wilson, G.E.; Zuber, N.

    1988-01-01

    Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs.

  18. Analysis of Maximum Reasonably Foreseeable Accidents for the Yucca Mountain Draft Environmental Impact Statement (DEIS)

    SciTech Connect

    S.B. Ross; R.E. Best; S.J. Maheras; T.I. McSweeney

    2001-08-17

    Accidents could occur during the transportation of spent nuclear fuel and high-level radioactive waste. This paper describes the risks and consequences to the public from accidents that are highly unlikely but that could have severe consequences. The impact of these accidents would include those to a collective population and to hypothetical maximally exposed individuals (MEIs). This document discusses accidents with conditions that have a chance of occurring more often than 1 in 10 million times in a year, called ''maximum reasonably foreseeable accidents''. Accidents and conditions less likely than this are not considered to be reasonably foreseeable.

  19. A systemic approach to accident analysis: a case study of the Stockwell shooting.

    PubMed

    Jenkins, Daniel P; Salmon, Paul M; Stanton, Neville A; Walker, Guy H

    2010-01-01

    This paper uses a systemic approach to accident investigation, based upon AcciMaps, to model the events leading up to the shooting of Jean Charles de Menezes at Stockwell Underground station in July 2005. The model captures many of the findings of the Independent Police Complaints Commission's report in a single representation, modelling their interdependencies and the causal flow. Furthermore, by taking a systemic approach, the analysis identifies further considerations related to the suitability of the Metropolitan Police Service's organisational structure to support rapid-paced operations, where reliable identification of a suspect is not possible. Based upon the analysis, the paper questions the division of functions between teams and the suitability of an organisational structure that relies upon the complex flow of information between separate teams for surveillance and for controlling the suspect. A dynamic organisational structure is proposed that changes in response to operation type and unfolding events. STATEMENT OF RELEVANCE: This paper provides much needed and called for validation for a systemic approach to accident analysis. A widely reported case study is used to illustrate the process. The paper shows how such an approach can consolidate the key findings of much larger reports as well as draw out additional recommendations. PMID:20069477

  20. Bicycle accidents often cause disability--an analysis of medical and social consequences of nonfatal bicycle accidents.

    PubMed

    Olkkonen, S; Lahdenranta, U; Slätis, P; Honkanen, R

    1993-06-01

    Social and medical consequences of 278 children and 264 adults injured in bicycle accidents and seen in two hospitals in Helsinki in 1985-86 were analyzed. Information was collected from patient records, by means of a special questionnaire and by telephone interview. A child outpatient required 1.7 and a child inpatient 3.0 physician visits on an average, while adults required 2.2 and 4.9 visits, respectively. The average duration of hospital stay was 8 days for hospitalized adults and 6 days for children. Rehabilitative care outside the hospital was received by 6% of the adult outpatients and 25% of the inpatients, but none of the injured children. The mean duration of work disability was 82 days among inpatients, 11 days among outpatients, 127 days among the inpatients injured in motor vehicle collisions and 65 days among inpatients injured in other bicycle accidents. Of inpatients 32% and of outpatients 5% reported persistent (> 6 months) disability. Persistent disability was recorded in 11% of children, in 47% of adults and in 67% of elderly inpatients. Most serious consequences were due to intracranial injuries in motor vehicle-bicycle collisions. Of the hospitalized bicyclists 4% suffered from severe cognitive and behavioural changes or sense impairment and of adult inpatients 3% suffered from permanent work disability. The average costs of health and social services were about FIM 1000 per adult outpatient and FIM 13000 per adult inpatient. In prevention high priority should be given to motor vehicle collisions, head injuries and injuries among the elderly bicyclists. PMID:8367689

  1. NASA Structural Analysis Report on the American Airlines Flight 587 Accident - Local Analysis of the Right Rear Lug

    NASA Technical Reports Server (NTRS)

    Raju, Ivatury S; Glaessgen, Edward H.; Mason, Brian H; Krishnamurthy, Thiagarajan; Davila, Carlos G

    2005-01-01

    A detailed finite element analysis of the right rear lug of the American Airlines Flight 587 - Airbus A300-600R was performed as part of the National Transportation Safety Board s failure investigation of the accident that occurred on November 12, 2001. The loads experienced by the right rear lug are evaluated using global models of the vertical tail, local models near the right rear lug, and a global-local analysis procedure. The right rear lug was analyzed using two modeling approaches. In the first approach, solid-shell type modeling is used, and in the second approach, layered-shell type modeling is used. The solid-shell and the layered-shell modeling approaches were used in progressive failure analyses (PFA) to determine the load, mode, and location of failure in the right rear lug under loading representative of an Airbus certification test conducted in 1985 (the 1985-certification test). Both analyses were in excellent agreement with each other on the predicted failure loads, failure mode, and location of failure. The solid-shell type modeling was then used to analyze both a subcomponent test conducted by Airbus in 2003 (the 2003-subcomponent test) and the accident condition. Excellent agreement was observed between the analyses and the observed failures in both cases. From the analyses conducted and presented in this paper, the following conclusions were drawn. The moment, Mx (moment about the fuselage longitudinal axis), has significant effect on the failure load of the lugs. Higher absolute values of Mx give lower failure loads. The predicted load, mode, and location of the failure of the 1985-certification test, 2003-subcomponent test, and the accident condition are in very good agreement. This agreement suggests that the 1985-certification and 2003- subcomponent tests represent the accident condition accurately. The failure mode of the right rear lug for the 1985-certification test, 2003-subcomponent test, and the accident load case is identified as a

  2. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    SciTech Connect

    Carbajo, J.J.

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  3. Quantification method analysis of the relationship between occupant injury and environmental factors in traffic accidents.

    PubMed

    Ju, Yong Han; Sohn, So Young

    2011-01-01

    Injury analysis following a vehicle crash is one of the most important research areas. However, most injury analyses have focused on one-dimensional injury variables, such as the AIS (Abbreviated Injury Scale) or the IIS (Injury Impairment Scale), at a time in relation to various traffic accident factors. However, these studies cannot reflect the various injury phenomena that appear simultaneously. In this paper, we apply quantification method II to the NASS (National Automotive Sampling System) CDS (Crashworthiness Data System) to find the relationship between the categorical injury phenomena, such as the injury scale, injury position, and injury type, and the various traffic accident condition factors, such as speed, collision direction, vehicle type, and seat position. Our empirical analysis indicated the importance of safety devices, such as restraint equipment and airbags. In addition, we found that narrow impact, ejection, air bag deployment, and higher speed are associated with more severe than minor injury to the thigh, ankle, and leg in terms of dislocation, abrasion, or laceration. PMID:21094332

  4. 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)

  5. Analysis of Occupational Accident Fatalities and Injuries Among Male Group in Iran Between 2008 and 2012

    PubMed Central

    Alizadeh, Seyed Shamseddin; Mortazavi, Seyed Bagher; Sepehri, Mohammad Mehdi

    2015-01-01

    Background: Because of occupational accidents, permanent disabilities and deaths occur and economic and workday losses emerge. Objectives: The purpose of the present study was to investigate the factors responsible for occupational accidents occurred in Iran. Patients and Methods: The current study analyzed 1464 occupational accidents recorded by the Ministry of Labor and Social Affairs’ offices in Iran during 2008 - 2012. At first, general understanding of accidents was obtained using descriptive statistics. Afterwards, the chi-square test and Cramer’s V statistic (Vc) were used to determine the association between factors influencing the type of injury as occupational accident outcomes. Results: There was no significant association between marital status and time of day with the type of injury. However, activity sector, cause of accident, victim’s education, age of victim and victim’s experience were significantly associated with the type of injury. Conclusions: Successful accident prevention relies largely on knowledge about the causes of accidents. In any accident control activity, particularly in occupational accidents, correctly identifying high-risk groups and factors influencing accidents is the key to successful interventions. Results of this study can cause to increase accident awareness and enable workplace’s management to select and prioritize problem areas and safety system weakness in workplaces. PMID:26568848

  6. A Content Analysis of News Media Coverage of the Accident at Three Mile Island.

    ERIC Educational Resources Information Center

    Stephens, Mitchell; Edison, Nadyne G.

    A study was conducted for the President's Commission on the Accident at Three Mile Island to analyze coverage of the accident by ten news organizations: two wire services, three commercial television networks, and five daily newspapers. Copies of all stories and transcripts of news programs during the first week of the accident were examined from…

  7. Effects of improved modeling on best estimate BWR severe accident analysis

    SciTech Connect

    Hyman, C.R.; Ott, L.J.

    1984-01-01

    Since 1981, ORNL has completed best estimate studies analyzing several dominant BWR accident scenarios. These scenarios were identified by early Probabilistic Risk Assessment (PRA) studies and detailed ORNL analysis complements such studies. In performing these studies, ORNL has used the MARCH code extensively. ORNL investigators have identified several deficiencies in early versions of MARCH with regard to BWR modeling. Some of these deficiencies appear to have been remedied by the most recent release of the code. It is the purpose of this paper to identify several of these deficiencies. All the information presented concerns the degraded core thermal/hydraulic analysis associated with each of the ORNL studies. This includes calculations of the containment response. The period of interest is from the time of permanent core uncovery to the end of the transient. Specific objectives include the determination of the extent of core damage and timing of major events (i.e., onset of Zr/H/sub 2/O reaction, initial clad/fuel melting, loss of control blade structure, etc.). As mentioned previously the major analysis tool used thus far was derived from an early version of MARCH. BWRs have unique features which must be modeled for best estimate severe accident analysis. ORNL has developed and incorporated into its version of MARCH several improved models. These include (1) channel boxes and control blades, (2) SRV actuations, (3) vessel water level, (4) multi-node analysis of in-vessel water inventory, (5) comprehensive hydrogen and water properties package, (6) first order correction to the ideal gas law, and (7) separation of fuel and cladding. Ongoing and future modeling efforts are required. These include (1) detailed modeling for the pressure suppression pool, (2) incorporation of B/sub 4/C/steam reaction models, (3) phenomenological model of corium mass transport, and (4) advanced corium/concrete interaction modeling. 10 references, 17 figures, 1 table.

  8. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    SciTech Connect

    Nathan V. Hoffer; Nolan A. Anderson; Piyush Sabharwall

    2011-08-01

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  9. Narrative text analysis of accident reports with tractors, self-propelled harvesting machinery and materials handling machinery in Austrian agriculture from 2008 to 2010 - a comparison.

    PubMed

    Mayrhofer, Hannes; Quendler, Elisabeth; Boxberger, Josef

    2014-01-01

    The aim of this study was the identification of accident scenarios and causes by analysing existing accident reports of recognized agricultural occupational accidents with tractors, self-propelled harvesting machinery and materials handling machinery from 2008 to 2010. As a result of a literature-based evaluation of past accident analyses, the narrative text analysis was chosen as an appropriate method. A narrative analysis of the text fields of accident reports that farmers used to report accidents to insurers was conducted to obtain detailed information about the scenarios and causes of accidents. This narrative analysis of reports was made the first time and yielded first insights for identifying antecedents of accidents and potential opportunities for technical based intervention. A literature and internet search was done to discuss and confirm the findings. The narrative text analysis showed that in more than one third of the accidents with tractors and materials handling machinery the vehicle rolled or tipped over. The most relevant accident scenarios with harvesting machinery were being trapped and falling down. The direct comparison of the analysed machinery categories showed that more than 10% of the accidents in each category were caused by technical faults, slippery or muddy terrain and incorrect or inappropriate operation of the vehicle. Accidents with tractors, harvesting machinery and materials handling machinery showed similarities in terms of causes, circumstances and consequences. Certain technical and communicative measures for accident prevention could be used for all three machinery categories. Nevertheless, some individual solutions for accident prevention, which suit each specific machine type, would be necessary. PMID:24738521

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

    SciTech Connect

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

    1986-09-01

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

  11. Analysis of station blackout accidents for the Bellefonte pressurized water reactor

    SciTech Connect

    Gasser, R D; Bieniarz, P P; Tills, J L

    1986-09-01

    An analysis has been performed for the Bellefonte PWR Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis which include the effects of direct heating on containment loading, and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating which involves more than about 50% of the core can fail the Bellefonte containment, but natural convection in the RCS may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach due to natural circulation and after vessel breach due to reevolution of retained fission products by fission product heating of RCS structures.

  12. Analysis of the SL-1 Accident Using RELAPS5-3D

    SciTech Connect

    Francisco, A.D. and Tomlinson, E. T.

    2007-11-08

    On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with a discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).

  13. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

    SciTech Connect

    Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harper, F.T.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

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

    SciTech Connect

    Wu, T

    2008-04-30

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

  15. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Harrison, J.D.; Harper, F.T.; Hora, S.C.

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  16. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

    SciTech Connect

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.; Boardman, J.; Jones, J.A.; Harper, F.T.; Young, M.L.; Hora, S.C.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  17. Learning from the Piper Alpha accident: A postmortem analysis of technical and organizational factors

    SciTech Connect

    Pate-Cornell, M.E. )

    1993-04-01

    The accident that occurred on board the offshore platform Piper Alpha in July 1988 killed 167 people and cost billions of dollars in property damage. It was caused by a massive fire, which was not the result of an unpredictable act of God' but of an accumulation of errors and questionable decisions. Most of them were rooted in the organization, its structure, procedures, and culture. This paper analyzes the accident scenario using the risk analysis framework, determines which human decision and actions influenced the occurrence of the basic events, and then identifies the organizational roots of these decisions and actions. These organizational factors are generalizable to other industries and engineering systems. They include flaws in the design guidelines and design practices (e.g., tight physical couplings or insufficient redundancies), misguided priorities in the management of the tradeoff between productivity and safety, mistakes in the management of the personnel on board, and errors of judgement in the process by which financial pressures are applied on the production sector (i.e., the oil companies' definition of profit centers) resulting in deficiencies in inspection and maintenance operations. This analytical approach allows identification of risk management measures that go beyond the purely technical (e.g., add redundancies to a safety system) and also include improvements of management practices. 18 refs., 4 figs.

  18. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

    SciTech Connect

    Haskin, F.E.; Harper, F.T.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  19. Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code

    SciTech Connect

    Chang Oh

    2007-05-01

    A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature. New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary). A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel and the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.

  20. A system analysis computer model for the High Flux Isotope Reactor (HFIRSYS Version 1)

    SciTech Connect

    Sozer, M.C.

    1992-04-01

    A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear`s stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system`s behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

  1. Development of a liquid jet model for implementation in a 3-dimensional Eularian analysis tool

    NASA Astrophysics Data System (ADS)

    Buschman, Francis X., III

    The ability to model the thermal behavior of a nuclear reactor is of utmost importance to the reactor designer. Condensation is an important phenomenon when modeling a reactor system's response to a Loss Of Coolant Accident (LOCA). Condensation is even more important with the use of passive safety systems which rely on condensation heat transfer for long term cooling. The increasing use of condensation heat transfer, including condensation on jets of water, in safety systems puts added pressure to correctly model this phenomenon with thermal-hydraulic system and sub-channel analysis codes. In this work, a stand alone module with which to simulate condensation on a liquid jet was developed and then implemented within a reactor vessel analysis code to improve that code's handling of jet condensation. It is shown that the developed liquid jet model vastly improves the ability of COBRA-TF to model condensation on turbulent liquid jets. The stand alone jet model and the coupled liquid jet COBRA-TF have been compared to experimental data. Jet condensation heat transfer experiments by Celata et al. with a variety of jet diameters, velocities, and subcooling were utilized to evaluate the models. A sensitivity study on the effects of noncondensables on jet condensation was also carried out using the stand alone jet model.

  2. Risk assessment of maintenance operations: the analysis of performing task and accident mechanism.

    PubMed

    Carrillo-Castrillo, Jesús A; Rubio-Romero, Juan Carlos; Guadix, Jose; Onieva, Luis

    2015-01-01

    Maintenance operations cover a great number of occupations. Most small and medium-sized enterprises lack the appropriate information to conduct risk assessments of maintenance operations. The objective of this research is to provide a method based on the concepts of task and accident mechanisms for an initial risk assessment by taking into consideration the prevalence and severity of the maintenance accidents reported. Data were gathered from 11,190 reported accidents in maintenance operations in the manufacturing sector of Andalusia from 2003 to 2012. By using a semi-quantitative methodology, likelihood and severity were evaluated based on the actual distribution of accident mechanisms in each of the tasks. Accident mechanisms and tasks were identified by using those variables included in the European Statistics of Accidents at Work methodology. As main results, the estimated risk of the most frequent accident mechanisms identified for each of the analysed tasks is low and the only accident mechanisms with medium risk are accidents when lifting or pushing with physical stress on the musculoskeletal system in tasks involving carrying, and impacts against objects after slipping or stumbling for tasks involving movements. The prioritisation of public preventive actions for the accident mechanisms with a higher estimated risk is highly recommended. PMID:25179119

  3. The Analysis of the Contribution of Human Factors to the In-Flight Loss of Control Accidents

    NASA Technical Reports Server (NTRS)

    Ancel, Ersin; Shih, Ann T.

    2012-01-01

    In-flight loss of control (LOC) is currently the leading cause of fatal accidents based on various commercial aircraft accident statistics. As the Next Generation Air Transportation System (NextGen) emerges, new contributing factors leading to LOC are anticipated. The NASA Aviation Safety Program (AvSP), along with other aviation agencies and communities are actively developing safety products to mitigate the LOC risk. This paper discusses the approach used to construct a generic integrated LOC accident framework (LOCAF) model based on a detailed review of LOC accidents over the past two decades. The LOCAF model is comprised of causal factors from the domain of human factors, aircraft system component failures, and atmospheric environment. The multiple interdependent causal factors are expressed in an Object-Oriented Bayesian belief network. In addition to predicting the likelihood of LOC accident occurrence, the system-level integrated LOCAF model is able to evaluate the impact of new safety technology products developed in AvSP. This provides valuable information to decision makers in strategizing NASA's aviation safety technology portfolio. The focus of this paper is on the analysis of human causal factors in the model, including the contributions from flight crew and maintenance workers. The Human Factors Analysis and Classification System (HFACS) taxonomy was used to develop human related causal factors. The preliminary results from the baseline LOCAF model are also presented.

  4. Modeling & analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    SciTech Connect

    Georgevich, V.; Kim, S.H.; Taleyarkhan, R.P.; Jackson, S.

    1992-10-01

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied.

  5. Chiropractic treatment of patients in motor vehicle accidents: a statistical analysis

    PubMed Central

    Dies, Stephen; Strapp, J Walter

    1992-01-01

    Motor vehicle accidents (MVA) are a major cause of spinal injuries treated by chiropractors. In this study the files of one chiropractor were reviewed retrospectively to generate a data base on the MVA cases (n = 149). The effect of age, sex, vehicle damage, symptoms and concurrent physiotherapy on the dependent variables of number of treatments, improvement and requirement for ongoing treatment was computed using an analysis of variance. Overall the average number of treatments given was 14.2. Patients who complained of headache or low back pain required more treatments than average. Improvement level was lowered by delay in seeking treatment, the presence of uncomplicated nausea and advancing age. Ongoing treatment to relieve persistent pain was required in 40.2 percent of the cases. None of the factors studied had a significant effect on this variable. The results of this study are comparable to those reported in the medical literature.

  6. Multifractal analysis of the 137Cs fallout pattern in Austria resulting from the Chernobyl accident.

    PubMed

    Pausch, G; Bossew, P; Hofmann, W; Steger, F

    1998-06-01

    The cumulative deposition of the 137Cs fallout in Austria resulting from the passage of the Chernobyl cloud has been investigated by applying correlation dimension and hyperbolic frequency distribution methods. For the analysis, a total of 1,881 deposition values were used, which were collected by the Federal Environmental Agency of Austria and the Federal Ministry of Health, representing all available measurements of 137Cs in soil made in Austria after the Chernobyl accident. From these data a hyperbolic exponent for the frequency distribution of 4.0 and a set of fractal correlation dimensions, which decrease from 1.426 +/- 0.022 (for the whole network) to 0.706 +/- 0.047 (for 137Cs values > or = 100 kBq m(-2)), were derived, thus confirming that the fallout pattern can be described as a multifractal. PMID:9600299

  7. PTSD Symptom Severity and Psychiatric Comorbidity in Recent Motor Vehicle Accident Victims: A Latent Class Analysis

    PubMed Central

    Hruska, Bryce; Irish, Leah A.; Pacella, Maria L.; Sledjeski, Eve M.; Delahanty, Douglas L.

    2014-01-01

    We conducted a latent class analysis (LCA) on 249 recent motor vehicle accident (MVA) victims to examine subgroups that differed in posttraumatic stress disorder (PTSD) symptom severity, current major depressive disorder and alcohol/other drug use disorders (MDD/AoDs), gender, and interpersonal trauma history 6-weeks post-MVA. A 4-class model best fit the data with a resilient class displaying asymptomatic PTSD symptom levels/low levels of comorbid disorders; a mild psychopathology class displaying mild PTSD symptom severity and current MDD; a moderate psychopathology class displaying severe PTSD symptom severity and current MDD/AoDs; and a severe psychopathology class displaying extreme PTSD symptom severity and current MDD. Classes also differed with respect to gender composition and history of interpersonal trauma experience. These findings may aid in the development of targeted interventions for recent MVA victims through the identification of subgroups distinguished by different patterns of psychiatric problems experienced 6-weeks post-MVA. PMID:25124501

  8. PTSD symptom severity and psychiatric comorbidity in recent motor vehicle accident victims: a latent class analysis.

    PubMed

    Hruska, Bryce; Irish, Leah A; Pacella, Maria L; Sledjeski, Eve M; Delahanty, Douglas L

    2014-10-01

    We conducted a latent class analysis (LCA) on 249 recent motor vehicle accident (MVA) victims to examine subgroups that differed in posttraumatic stress disorder (PTSD) symptom severity, current major depressive disorder and alcohol/other drug use disorders (MDD/AoDs), gender, and interpersonal trauma history 6-weeks post-MVA. A 4-class model best fit the data with a resilient class displaying asymptomatic PTSD symptom levels/low levels of comorbid disorders; a mild psychopathology class displaying mild PTSD symptom severity and current MDD; a moderate psychopathology class displaying severe PTSD symptom severity and current MDD/AoDs; and a severe psychopathology class displaying extreme PTSD symptom severity and current MDD. Classes also differed with respect to gender composition and history of interpersonal trauma experience. These findings may aid in the development of targeted interventions for recent MVA victims through the identification of subgroups distinguished by different patterns of psychiatric problems experienced 6-weeks post-MVA. PMID:25124501

  9. 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.

  10. Final safety analysis report for the Galileo Mission: Volume 2: Book 1, Accident model document

    SciTech Connect

    Not Available

    1988-12-15

    The Accident Model Document (AMD) is the second volume of the three volume Final Safety Analysis Report (FSAR) for the Galileo outer planetary space science mission. This mission employs Radioisotope Thermoelectric Generators (RTGs) as the prime electrical power sources for the spacecraft. Galileo will be launched into Earth orbit using the Space Shuttle and will use the Inertial Upper Stage (IUS) booster to place the spacecraft into an Earth escape trajectory. The RTG's employ silicon-germanium thermoelectric couples to produce electricity from the heat energy that results from the decay of the radioisotope fuel, Plutonium-238, used in the RTG heat source. The heat source configuration used in the RTG's is termed General Purpose Heat Source (GPHS), and the RTG's are designated GPHS-RTGs. The use of radioactive material in these missions necessitates evaluations of the radiological risks that may be encountered by launch complex personnel as well as by the Earth's general population resulting from postulated malfunctions or failures occurring in the mission operations. The FSAR presents the results of a rigorous safety assessment, including substantial analyses and testing, of the launch and deployment of the RTGs for the Galileo mission. This AMD is a summary of the potential accident and failure sequences which might result in fuel release, the analysis and testing methods employed, and the predicted source terms. Each source term consists of a quantity of fuel released, the location of release and the physical characteristics of the fuel released. Each source term has an associated probability of occurrence. 27 figs., 11 tabs.

  11. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project.

  12. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

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

    SciTech Connect

    Burgherr, P.; Hirschberg, S.

    2008-07-01

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

  14. Analysis of the Uniform Accident And Sickness Policy Provision Law: lessons for social work practice, policy, and research.

    PubMed

    Cochran, Gerald

    2010-01-01

    The Uniform Accident and Sickness Policy Provision Law (UPPL) is a state statute that allows insurance companies in 26 states to deny claims for accidents and injuries incurred by persons under the influence of drugs or alcohol. Serious repercussions can result for patients and health care professionals as states enforce this law. To examine differences within the laws that might facilitate amendments or reduce insurance companies' ability to deny claims, a content analysis was carried out of each state's UPPL law. Results showed no meaningful differences between each state's laws. These results indicate patients and health professionals share similar risk related to the UPPL regardless of state. PMID:20711944

  15. Defense In-Depth Accident Analysis Evaluation of Tritium Facility Bldgs. 232-H, 233-H, and 234-H

    SciTech Connect

    Blanchard, A.

    1999-05-10

    'The primary purpose of this report is to document a Defense-in-Depth (DID) accident analysis evaluation for Department of Energy (DOE) Savannah River Site (SRS) Tritium Facility Buildings 232-H, 233-H, and 234-H. The purpose of a DID evaluation is to provide a more realistic view of facility radiological risks to the offsite public than the bounding deterministic analysis documented in the Safety Analysis Report, which credits only Safety Class items in the offsite dose evaluation.'

  16. Posttest analysis of MIST Test 3109AA using TRAC-PF1/MOD1

    SciTech Connect

    Steiner, J.L.; Siebe, D.A.; Boyack, B.E. )

    1992-09-01

    This document discusses a posttest calculation and analysis of Multi-loop Integral System Test (MIST) 3109AA as the nominal test for the MIST program. It is a test of a small-break loss-of-coolant accident (SBLOCA) with a scaled 10-cm[sup 2] break in the B1 cold leg. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, intermittent and interrupted loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator secondaries was used after steam-generator secondary refill and symmetric steam-generator pressure control was used. We performed the calculation using TRAC-PF1/MODI. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. It is believed that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.

  17. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  18. 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.

  19. The Wheels of Misfortune: A Time Series Analysis of Bicycle Accidents on a College Campus.

    ERIC Educational Resources Information Center

    Johnson, Mark S.; And Others

    1978-01-01

    The effectiveness of engineering and policy interventions in reducing bicycle accidents on the campus of the University of California at Santa Barbara was investigated. None of the bikeway modifications were found to have been effective in reducing bicycle accidents. (Author/GDC)

  20. 3W approach to the investigation, analysis, and prevention of human-error aircraft accidents.

    PubMed

    Ricketson, D S; Brown, W R; Graham, K N

    1980-09-01

    Human error is the largest cause of U.S. Army aircraft accidents. An approach to this problem is presented which is based on a model of the human-error accident. This 3W approach identifies what task error (TE) caused or contributed to the accident, what inadequacy (I) in the aviation system caused or allowed the TE to occur, and what remedial measure (R) is required to correct the I. There were 82 human-error accidents analyzed to identify TEIR information. Statistically important is were identified which could be remedied based on accident costs. Then, potentially cost-effective remedial actions were ranked on a cost-benefit totem pole. The totem pole was given to the aviation system manager as a management tool to assist in determining priorities for corrective actions. PMID:7417175

  1. Emergency drinking water treatment during source water pollution accidents in China: origin analysis, framework and technologies.

    PubMed

    Zhang, Xiao-Jian; Chen, Chao; Lin, Peng-Fei; Hou, Ai-Xin; Niu, Zhang-Bin; Wang, Jun

    2011-01-01

    China has suffered frequent source water contamination accidents in the past decade, which has resulted in severe consequences to the water supply of millions of residents. The origins of typical cases of contamination are discussed in this paper as well as the emergency response to these accidents. In general, excessive pursuit of rapid industrialization and the unreasonable location of factories are responsible for the increasing frequency of accidental pollution events. Moreover, insufficient attention to environmental protection and rudimentary emergency response capability has exacerbated the consequences of such accidents. These environmental accidents triggered or accelerated the promulgation of stricter environmental protection policy and the shift from economic development mode to a more sustainable direction, which should be regarded as the turning point of environmental protection in China. To guarantee water security, China is trying to establish a rapid and effective emergency response framework, build up the capability of early accident detection, and develop efficient technologies to remove contaminants from water. PMID:21133359

  2. Hazard categorization and accident analysis techniques for compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports

    SciTech Connect

    1992-12-31

    The purpose of this DOE Standard is to establish guidance for facility managers and Program Secretarial Officers (PSOs) and thereby help them to comply consistently and more efficiently with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports. To this end, this guidance provides the following practical information: (1) The threshold quantities of radiological material inventory below which compliance with DOE Order 5480.23 is not required. (2) The level of effort to develop the program plan and schedule required in Section 9.b. (2) of the Order, and information for making a preliminary assessment of facility hazards. (3) A uniform methodology for hazard categorization under the Order. (4) Insight into the ''graded approach'' for SAR development, especially in hazard assessment and accident analysis techniques. Individual PSOs may develop additional guidance addressing safety requirements for facilities which fall below the threshold quantities specified in this document.

  3. Investigation of Countercurrent Helium-Air Flows in Air-ingress Accidents for VHTRs

    SciTech Connect

    Sun, Xiaodong; Christensen, Richard; Oh, Chang

    2013-10-03

    The primary objective of this research is to develop an extensive experimental database for the air- ingress phenomenon for the validation of computational fluid dynamics (CFD) analyses. This research is intended to be a separate-effects experimental study. However, the project team will perform a careful scaling analysis prior to designing a scaled-down test facility in order to closely tie this research with the real application. As a reference design in this study, the team will use the 600 MWth gas turbine modular helium reactor (GT-MHR) developed by General Atomic. In the test matrix of the experiments, researchers will vary the temperature and pressure of the helium— along with break size, location, shape, and orientation—to simulate deferent scenarios and to identify potential mitigation strategies. Under support of the Department of Energy, a high-temperature helium test facility has been designed and is currently being constructed at Ohio State University, primarily for high- temperature compact heat exchanger testing for the VHTR program. Once the facility is in operation (expected April 2009), this study will utilize high-temperature helium up to 900°C and 3 MPa for loss-of-coolant accident (LOCA) depressurization and air-ingress experiments. The project team will first conduct a scaling study and then design an air-ingress test facility. The major parameter to be measured in the experiments is oxygen (or nitrogen) concentration history at various locations following a LOCA scenario. The team will use two measurement techniques: 1) oxygen (or similar type) sensors employed in the flow field, which will introduce some undesirable intrusiveness, disturbing the flow, and 2) a planar laser-induced fluorescence (PLIF) imaging technique, which has no physical intrusiveness to the flow but requires a transparent window or test section that the laser beam can penetrate. The team will construct two test facilities, one for high-temperature helium tests with

  4. 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

  5. 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

  6. Development of NASA's Accident Precursor Analysis Process Through Application on the Space Shuttle Orbiter

    NASA Technical Reports Server (NTRS)

    Maggio, Gaspare; Groen, Frank; Hamlin, Teri; Youngblood, Robert

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system. APA docs more than simply track experience: it systematically evaluates experience, looking for under-appreciated risks that may warrant changes to design or operational practice. This paper presents the pilot application of the NASA APA process to Space Shuttle Orbiter systems. In this effort, the working sessions conducted at Johnson Space Center (JSC) piloted the APA process developed by Information Systems Laboratories (ISL) over the last two years under the auspices of NASA's Office of Safety & Mission Assurance, with the assistance of the Safety & Mission Assurance (S&MA) Shuttle & Exploration Analysis Branch. This process is built around facilitated working sessions involving diverse system experts. One important aspect of this particular APA process is its focus on understanding the physical mechanism responsible for an operational anomaly, followed by evaluation of the risk significance of the observed anomaly as well as consideration of generalizations of the underlying mechanism to other contexts. Model completeness will probably always be an issue, but this process tries to leverage operating experience to the extent possible in order to address completeness issues before a catastrophe occurs.

  7. Overview of the Aerothermodynamics Analysis Conducted in Support of the STS-107 Accident Investigation

    NASA Technical Reports Server (NTRS)

    Campbell, Charles H.

    2004-01-01

    A graphic presentation of the aerothermodynamics analysis conducted in support of the STS-107 accident investigation. Investigation efforts were conducted as part of an integrated AATS team (Aero, Aerothermal, Thermal, Stress) directed by OVEWG. Graphics presented are: STS-107 Entry trajectory and timeline (1st off-nominal event to Post-LOS); Indications from OI telemetry data; Aero/aerothermo/thermal analysis process; Selected STS-107 side fuselage/OMS pod off-nominal temperatures; Leading edge structural subsystem; Relevant forensics evidence; External aerothermal environments; STS-107 Pre-entry EOM3 heating profile; Surface heating and temperatures; Orbiter wing leading edge damage survey; Internal aerothermal environments; Orbiter wing CAD model; Aerodynamic flight reconstruction; Chronology of aerodynamic/aerothermoydynamic contributions; Acreage TPS tile damage; Larger OML perturbations; Missing RCC panel(s); Localized damage to RCC panel/missing T-seal; RCC breach with flow ingestion; and Aero-aerothermal closure. NAIT served as the interface between the CAIB and NASA investigation teams; and CAIB requests for study were addressed.

  8. The role of mitochondrial proteomic analysis in radiological accidents and terrorism.

    PubMed

    Maguire, David; Zhang, Bingrong; Zhang, Amy; Zhang, Lurong; Okunieff, Paul

    2013-01-01

    In the wake of the 9/11 terrorist attacks and the recent Level 7 nuclear event at the Fukushima Daiichi plant, there has been heightened awareness of the possibility of radiological terrorism and accidents and the need for techniques to estimate radiation levels after such events. A number of approaches to monitoring radiation using biological markers have been published, including physical techniques, cytogenetic approaches, and direct, DNA-analysis approaches. Each approach has the potential to provide information that may be applied to the triage of an exposed population, but problems with development and application of devices or lengthy analyses limit their potential for widespread application. We present a post-irradiation observation with the potential for development into a rapid point-of-care device. Using simple mitochondrial proteomic analysis, we investigated irradiated and nonirradiated murine mitochondria and identified a protein mobility shift occurring at 2-3 Gy. We discuss the implications of this finding both in terms of possible mechanisms and potential applications in bio-radiation monitoring. PMID:22879026

  9. LOCA simulation: analysis of rarefaction waves propagating through geometric singularities

    SciTech Connect

    Crouzet, Fabien; Faucher, Vincent; Galon, Pascal; Piteau, Philippe; Izquierdo, Patrick

    2012-07-01

    The propagation of a transient wave through an orifice is investigated for applications to Loss Of Coolant Accident in nuclear plants. An analytical model is proposed for the response of an orifice plate and implemented in the EUROPLEXUS fast transient dynamics software. It includes an acoustic inertial effect in addition to a quasi-steady dissipation term. The model is experimentally validated on a test rig consisting in a single pipe filled with pressurized water. The test rig is designed to generate a rapid depressurization of the pipe, by means of a bursting disk. The proposed model gives results which compare favourably with experimental data. (authors)

  10. Analysis of offsite Emergency Planning Zones (EPZs) for the Rocky Flats Plant. Phase 3, Sitewide spectrum-of-accidents and bounding EPZ analysis

    SciTech Connect

    Petrocchi, A.J.; Zimmerman, G.A.

    1994-03-14

    During Phase 3 of the EPZ project, a sitewide analysis will be performed applying a spectrum-of-accidents approach to both radiological and nonradiological hazardous materials release scenarios. This analysis will include the MCA but will be wider in scope and will produce options for the State of Colorado for establishing a bounding EPZ that is intended to more comprehensively update the interim, preliminary EPZ developed in Phase 2. EG&G will propose use of a hazards assessment methodology that is consistent with the DOE Emergency Management Guide for Hazards Assessments and other methods required by DOE orders. This will include hazards, accident, safety, and risk analyses. Using this methodology, EG&G will develop technical analyses for a spectrum of accidents. The analyses will show the potential effects from the spectrum of accidents on the offsite population together with identification of offsite vulnerable zones and areas of concern. These analyses will incorporate state-of-the-art technology for accident analysis, atmospheric plume dispersion modeling, consequence analysis, and the application of these evaluations to the general public population at risk. The analyses will treat both radiological and nonradiological hazardous materials and mixtures of both released accidentally to the atmosphere. DOE/RFO will submit these results to the State of Colorado for the State`s use in determining offsite emergency planning zones for the Rocky Flats Plant. In addition, the results will be used for internal Rocky Flats Plant emergency planning.

  11. Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management

    NASA Astrophysics Data System (ADS)

    Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

    2016-04-01

    Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8 % of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced.

  12. Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management.

    PubMed

    Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

    2016-04-01

    Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8% of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced. PMID:26739714

  13. An analysis of civil aviation propeller-to-person accidents: 1965-79.

    PubMed

    Collins, W E; Mastrullo, A R; Kirkham, W R; Taylor, D K; Grape, P M

    1982-05-01

    The interest of manufacturing, governmental, and safety personnel in using paint schemes on propeller and rotor blades is based on improving the visual conspicuity of those blades when they are rotating. While propeller and rotor paint schemes may serve to reduce the number of fatalities and injuries due to contact with a rotating blade, there is little information about the circumstances surrounding such accidents. Brief reports provided by the National Transportation Safety Board of all "propeller-to-person" accidents from 1965-79 were examined and analyzed in terms of airport lighting conditions, actions of pilots, actions of passengers and groundcrew, phase of flight operations, weather conditions, and others. Analyses based on 319 accidents showed a marked drop in the frequency of "propeller-to-person" accidents from 1975 through 1978. Several types of educational efforts directed toward pilots and groundcrew, both prior to and during that 4-year period, were examined as possible factors contributing to the accident rate decline. Accident patterns provide a basis for assessing the probable efficacy of various recommendations, including propeller conspicuity, for further reducing "propeller-to-person" accidents. PMID:7092754

  14. Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

    NASA Astrophysics Data System (ADS)

    Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

    2014-05-01

    The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

  15. The effect of gamma-ray transport on afterheat calculations for accident analysis

    SciTech Connect

    Reyes, S.; Latkowski, J.F.; Sanz, J.

    2000-05-01

    Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed.

  16. The Role of Materials Degradation and Analysis in the Space Shuttle Columbia Accident Investigation

    NASA Technical Reports Server (NTRS)

    McDanels, Steven J.

    2006-01-01

    The efforts following the loss of the Space Shuttle Columbia included debris recovery, reconstruction, and analysis. The debris was subjected to myriad quantitative and semiquantitative chemical analysis techniques, ranging from examination via the scanning electron microscope (SEM) with energy dispersive spectrometer (EDS) to X-Ray diffraction (XRD) and electron probe micro-analysis (EPMA). The results from the work with the debris helped the investigators determine the location where a breach likely occurred in the leading edge of the left wing during lift off of the Orbiter from the Kennedy Space Center. Likewise, the information evidenced by the debris was also crucial in ascertaining the path of impinging plasma flow once it had breached the wing. After the Columbia Accident Investigation Board (CAIB) issued its findings, the major portion of the investigation was concluded. However, additional work remained to be done on many pieces of debris from portions of the Orbiter which were not directly related to the initial impact during ascent. This subsequent work was not only performed in the laboratory, but was also performed with portable equipment, including examination via portable X-Ray fluorescence (XRF) and Fourier transform infrared spectroscopy (FTIR). Likewise, acetate and silicon-rubber replicas of various fracture surfaces were obtained for later macroscopic and fractographic examination. This paper will detail the efforts and findings from the initial investigation, as well as present results obtained by the later examination and analysis of debris from the Orbiter including its windows, bulkhead structures, and other components which had not been examined during the primary investigation.

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

    SciTech Connect

    Sang Lung, Chan

    2004-07-01

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

  18. An analysis of the consequences of accidents involving shipments of multiple Type A radioactive material (RAM) packages

    SciTech Connect

    Finley, N.C.; McClure, J.D.; Reardon, P.C.; Wangler, M.

    1989-01-01

    Comparing the results of the RADTRANIII calculations with a normalized set of results, both for incident-free transport and vehicular accident cases, the calculated consequences in the current analysis are lower. Even for the High-Activity Shipment, the total expected population dose from either incident-free transport or vehicular accidents is small, and smaller than that estimated in USNRC 1977. The results of the simulation in which parameters were varied randomly and independently indicate that, regardless of the input values assumed, the maximum total population dose from the High-Activity Shipment and the simultaneous occurrence of the least conservative value for each input parameter might be as high as 300 person-rem for a single shipment. The values for either of the other shipments (DOT Exemption or Common Carrier) would be significantly lower. The potential average individual radiation doses from accidents involving multiple Type A package shipments are comparable to the increase in the normal background radiation dose of 0.09 rem/person/year (90 mrem) that an individual would receive by moving from sea level to 5000 ft elevation. The maximum dose to an individual (one very near the accident scene) for the High Activity Shipment would be approximately 0.3 rem (300 mrem) in a maximum severity accident. This is within the individual dose guidelines outlined by NCRP (0.5 rem). Even at the high levels postulated for multiple package shipments under DOT controlled exemptions, the potential risks to the public in terms of expected population dose in the current analysis are below those already found to be acceptable. 4 refs., 3 tabs.

  19. Analysis of labour accidents in tunnel construction and introduction of prevention measures.

    PubMed

    Kikkawa, Naotaka; Itoh, Kazuya; Hori, Tomohito; Toyosawa, Yasuo; Orense, Rolando P

    2015-01-01

    At present, almost all mountain tunnels in Japan are excavated and constructed utilizing the New Austrian Tunneling Method (NATM), which was advocated by Prof. Rabcewicz of Austria in 1964. In Japan, this method has been applied to tunnel construction since around 1978, after which there has been a subsequent decrease in the number of casualties during tunnel construction. However, there is still a relatively high incidence of labour accidents during tunnel construction when compared to incidence rates in the construction industry in general. During tunnel construction, rock fall events at the cutting face are a particularly characteristic of the type of accident that occurs. In this study, we analysed labour accidents that possess the characteristics of a rock fall event at a work site. We also introduced accident prevention measures against rock fall events. PMID:26027707

  20. Source terms for analysis of accidents at a high level waste repository

    SciTech Connect

    Mubayi, V.; Davis, R.E.; Youngblood, R.

    1989-01-01

    This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs.

  1. Traffic Analysis and Road Accidents: A Case Study of Hyderabad using GIS

    NASA Astrophysics Data System (ADS)

    Bhagyaiah, M.; Shrinagesh, B.

    2014-06-01

    Globalization has impacted many developing countries across the world. India is one such country, which benefited the most. Increased, economic activity raised the consumption levels of the people across the country. This created scope for increase in travel and transportation. The increase in the vehicles since last 10 years has put lot of pressure on the existing roads and ultimately resulting in road accidents. It is estimated that since 2001 there is an increase of 202 percent of two wheeler and 286 percent of four wheeler vehicles with no road expansion. Motor vehicle crashes are a common cause of death, disability and demand for emergency medical care. Globally, more than 1 million people die each year from traffic crashes and about 20-50 million are injured or permanently disabled. There has been increasing trend in road accidents in Hyderabad over a few years. GIS helps in locating the accident hotspots and also in analyzing the trend of road accidents in Hyderabad.

  2. Analysis of labour accidents in tunnel construction and introduction of prevention measures

    PubMed Central

    KIKKAWA, Naotaka; ITOH, Kazuya; HORI, Tomohito; TOYOSAWA, Yasuo; ORENSE, Rolando P.

    2015-01-01

    At present, almost all mountain tunnels in Japan are excavated and constructed utilizing the New Austrian Tunneling Method (NATM), which was advocated by Prof. Rabcewicz of Austria in 1964. In Japan, this method has been applied to tunnel construction since around 1978, after which there has been a subsequent decrease in the number of casualties during tunnel construction. However, there is still a relatively high incidence of labour accidents during tunnel construction when compared to incidence rates in the construction industry in general. During tunnel construction, rock fall events at the cutting face are a particularly characteristic of the type of accident that occurs. In this study, we analysed labour accidents that possess the characteristics of a rock fall event at a work site. We also introduced accident prevention measures against rock fall events. PMID:26027707

  3. Sensitivity analysis of a ship accident at a deep-ocean site in the northwest Atlantic

    SciTech Connect

    Kaplan, M.F.

    1985-04-01

    This report presents the results of a sensitivity analysis for an HLW ship accident occurring in the Nares Abyssal Plain in the northwestern Atlantic. Waste form release rate, canister lifetime and sorption in the water column (partition coefficients) were varied. Also investigated were the relative importance of the dose from the food chain and from seaweed in the diet. Peak individual doses and integrated collective doses for populations were the units of comparison. In accordance with international guidelines on radiological protection, the comparisons of different options were carried out over ''all time''; the study uses a million-year time frame. Partition coefficients have the most pronounced effect on collective dose of the parameters studied. Variations in partition coefficients affect the shape of the collective dose curve over the entire time frame. Peak individual doses decrease markedly when the value for the sorption of americium is increased, but show no increase when less sorption is assumed. Waste form release rates and canister lifetimes affect collective doses only in periods prior to 20,000 years. Hence, comparisons of these options need not be carried out beyond 20,000 years. Waste from release rates below 10/sup -3//yr (nominal value) affect individual doses in a linear manner, i.e., an order-of-magnitude reduction in release rate leads to an order-of-magnitude reduction in peak individual dose. Little reduction in peak individual doses is seen with canister lifetimes extended beyond the nominal 100 years. 32 refs., 14 figs., 16 tabs.

  4. THERMAL ANALYSIS OF A 9975 PACKAGE IN A FACILITY FIRE ACCIDENT

    SciTech Connect

    Gupta, N.

    2011-02-14

    Surplus plutonium bearing materials in the U.S. Department of Energy (DOE) complex are stored in the 3013 containers that are designed to meet the requirements of the DOE standard DOE-STD-3013. The 3013 containers are in turn packaged inside 9975 packages that are designed to meet the NRC 10 CFR Part 71 regulatory requirements for transporting the Type B fissile materials across the DOE complex. The design requirements for the hypothetical accident conditions (HAC) involving a fire are given in 10 CFR 71.73. The 9975 packages are stored at the DOE Savannah River Site in the K-Area Material Storage (KAMS) facility for long term of up to 50 years. The design requirements for safe storage in KAMS facility containing multiple sources of combustible materials are far more challenging than the HAC requirements in 10 CFR 71.73. While the 10 CFR 71.73 postulates an HAC fire of 1475 F and 30 minutes duration, the facility fire calls for a fire of 1500 F and 86 duration. This paper describes a methodology and the analysis results that meet the design limits of the 9975 component and demonstrate the robustness of the 9975 package.

  5. Identification of Behavior Based Safety by Using Traffic Light Analysis to Reduce Accidents

    NASA Astrophysics Data System (ADS)

    Mansur, A.; Nasution, M. I.

    2016-01-01

    This work present the safety assessment of a case study and describes an important area within the field production in oil and gas industry, namely behavior based safety (BBS). The company set a rigorous BBS and its intervention program that implemented and deployed continually. In this case, observers requested to have discussion and spread a number of determined questions related with work behavior to the workers during observation. Appraisal of Traffic Light Analysis (TLA) as one tools of risk assessment used to determine the estimated score of BBS questionnaire. Standardization of TLA appraisal in this study are based on Regulation of Minister of Labor and Occupational Safety and Health No:PER.05/MEN/1996. The result shown that there are some points under 84%, which categorized in yellow category and should corrected immediately by company to prevent existing bad behavior of workers. The application of BBS expected to increase the safety performance at work time-by-time and effective in reducing accidents.

  6. 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.

  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. 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

  9. Highway accident severities and the mixed logit model: an exploratory empirical analysis.

    PubMed

    Milton, John C; Shankar, Venky N; Mannering, Fred L

    2008-01-01

    Many transportation agencies use accident frequencies, and statistical models of accidents frequencies, as a basis for prioritizing highway safety improvements. However, the use of accident severities in safety programming has been often been limited to the locational assessment of accident fatalities, with little or no emphasis being placed on the full severity distribution of accidents (property damage only, possible injury, injury)-which is needed to fully assess the benefits of competing safety-improvement projects. In this paper we demonstrate a modeling approach that can be used to better understand the injury-severity distributions of accidents on highway segments, and the effect that traffic, highway and weather characteristics have on these distributions. The approach we use allows for the possibility that estimated model parameters can vary randomly across roadway segments to account for unobserved effects potentially relating to roadway characteristics, environmental factors, and driver behavior. Using highway-injury data from Washington State, a mixed (random parameters) logit model is estimated. Estimation findings indicate that volume-related variables such as average daily traffic per lane, average daily truck traffic, truck percentage, interchanges per mile and weather effects such as snowfall are best modeled as random-parameters-while roadway characteristics such as the number of horizontal curves, number of grade breaks per mile and pavement friction are best modeled as fixed parameters. Our results show that the mixed logit model has considerable promise as a methodological tool in highway safety programming. PMID:18215557

  10. Biomechanical analysis of occupant kinematics in rollover motor vehicle accidents: dynamic spit test.

    PubMed

    Sances, Anthony; Kumaresan, Srirangam; Clarke, Richard; Herbst, Brian; Meyer, Steve

    2005-01-01

    A better understanding of occupant kinematics in rollover accidents helps to advance biomechanical knowledge and to enhance the safety features of motor vehicles. While many rollover accident simulation studies have adopted the static approach to delineate the occupant kinematics in rollover accidents, very few studies have attempted the dynamic approach. The present work was designed to study the biomechanics of restrained occupants during rollover accidents using the steady-state dynamic spit test and to address the importance of keeping the lap belt fastened. Experimental tests were conducted using an anthropometric 50% Hybrid III dummy in a vehicle. The vehicle was rotated at 180 degrees/second and the dummy was restrained using a standard three-point restraint system. The lap belt of the dummy was fastened either by using the cinching latch plate or by locking the retractor. Three configurations of shoulder belt harness were simulated: shoulder belt loose on chest with cinch plate, shoulder belt under the left arm and shoulder belt behind the chest. In all tests, the dummy stayed within the confinement of the vehicle indicating that the securely fastened lap belt holds the dummy with dynamic movement of 3 1/2" to 4". The results show that occupant movement in rollover accidents is least affected by various shoulder harness positions with a securely fastened lap belt. The present study forms a first step in delineating the biomechanics of occupants in rollover accidents. PMID:15850090

  11. Nasal continuous positive airway pressure (nCPAP) treatment for obstructive sleep apnea, road traffic accidents and driving simulator performance: a meta-analysis.

    PubMed

    Antonopoulos, Constantine N; Sergentanis, Theodoros N; Daskalopoulou, Styliani S; Petridou, Eleni Th

    2011-10-01

    We used meta-analysis to synthesize current evidence regarding the effect of nasal continuous positive airway pressure (nCPAP) on road traffic accidents in patients with obstructive sleep apnea (OSA) as well as on their performance in driving simulator. The primary outcomes were real accidents, near miss accidents, and accident-related events in the driving simulator. Pooled odds ratios (ORs), incidence rate ratios (IRRs) and standardized mean differences (SMDs) were appropriately calculated through fixed or random effects models after assessing between-study heterogeneity. Furthermore, risk differences (RDs) and numbers needed to treat (NNTs) were estimated for real and near miss accidents. Meta-regression analysis was performed to examine the effect of moderator variables and publication bias was also evaluated. Ten studies on real accidents (1221 patients), five studies on near miss accidents (769 patients) and six studies on the performance in driving simulator (110 patients) were included. A statistically significant reduction in real accidents (OR=0.21, 95% CI=0.12-0.35, random effects model; IRR=0.45, 95% CI=0.34-0.59, fixed effects model) and near miss accidents (OR=0.09, 95% CI=0.04-0.21, random effects model; IRR=0.23, 95% CI=0.08-0.67, random effects model) was observed. Likewise, a significant reduction in accident-related events was observed in the driving simulator (SMD=-1.20, 95% CI=-1.75 to -0.64, random effects). The RD for real accidents was -0.22 (95% CI=-0.32 to -0.13, random effects), with NNT equal to five patients (95% CI=3-8), whereas for near miss accidents the RD was -0.47 (95% CI=-0.69 to -0.25, random effects), with NNT equal to two patients (95% CI=1-4). For near miss accidents, meta-regression analysis suggested that nCPAP seemed more effective among patients entering the studies with higher baseline accident rates. In conclusion, all three meta-analyses demonstrated a sizeable protective effect of nCPAP on road traffic accidents, both

  12. Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities; Yucca Mountain Site Characterization Project

    SciTech Connect

    Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J.; Laub, T.W.

    1992-06-01

    This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10{sup {minus}11}/yr to 10{sup {minus}5}/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10{sup {minus}9}/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution.

  13. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    SciTech Connect

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  14. Timing analysis of PWR fuel pin failures

    SciTech Connect

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. ); Straka, M. )

    1992-09-01

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively.

  15. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    SciTech Connect

    Banati, J.

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  16. Retrospection of Chernobyl nuclear accident for decision analysis concerning remedial actions in Ukraine

    SciTech Connect

    Georgievskiy, Vladimir

    2007-07-01

    It is considered the efficacy of decisions concerning remedial actions when of-site radiological monitoring in the early and (or) in the intermediate phases was absent or was not informative. There are examples of such situations in the former Soviet Union where many people have been exposed: releases of radioactive materials from 'Krasnoyarsk-26' into Enisey River, releases of radioactive materials from 'Chelabinsk-65' (the Kishtim accident), nuclear tests at the Semipalatinsk Test Site, the Chernobyl nuclear accident etc. If monitoring in the early and (or) in the intermediate phases is absent the decisions concerning remedial actions are usually developed on the base of permanent monitoring. However decisions of this kind may be essentially erroneous. For these cases it is proposed to make retrospection of radiological data of the early and intermediate phases of nuclear accident and to project decisions concerning remedial actions on the base of both retrospective data and permanent monitoring data. In this Report the indicated problem is considered by the example of the Chernobyl accident for Ukraine. Their of-site radiological monitoring in the early and intermediate phases was unsatisfactory. In particular, the pasture-cow-milk monitoring had not been made. All official decisions concerning dose estimations had been made on the base of measurements of {sup 137}Cs in body (40 measurements in 135 days and 55 measurements in 229 days after the Chernobyl accident). For the retrospection of radiological data of the Chernobyl accident dynamic model has been developed. This model has structure similar to the structure of Pathway model and Farmland model. Parameters of the developed model have been identified for agricultural conditions of Russia and Ukraine. By means of this model dynamics of 20 radionuclides in pathways and dynamics of doses have been estimated for the early, intermediate and late phases of the Chernobyl accident. The main results are following

  17. Testing and analysis of structural integrity of electrosleeved tubes under severe accident transients

    SciTech Connect

    Majumdar, S.

    1999-12-10

    The structural integrity of flawed steam generator tubing with Electrosleeves{trademark} under simulated severe accident transients was analyzed by analytical models that used available material properties data and results from high-temperature tests conducted on Electrosleeved tubes. The Electrosleeve material is almost pure Ni and derives its strength and other useful properties from its nanocrystalline microstructure, which is stable at reactor operating temperatures. However, it undergoes rapid grain growth, at the high temperatures expected during severe accidents, resulting in a loss of strength and a corresponding decrease in flow stress. The magnitude of this decrease depends on the time-temperature history during the accident. Failure tests were conducted at ANL and FTI on internally pressurized Electrosleeved tubes with 80% and 100% throughwall machined axial notches in tie parent tubes that were subjected to simulated severe accident temperature transients. The test results, together with the analytical model, were used to estimate the unaged flow stress curve of the Electrosleeved material at high temperatures. Failure temperatures for Electrosleeved tubes with throughwall and part-throughwall axial cracks of various lengths in the parent tubes were calculated for a postulated severe accident transient.

  18. Accident analysis of large-scale technological disasters applied to an anaesthetic complication.

    PubMed

    Eagle, C J; Davies, J M; Reason, J

    1992-02-01

    The occurrence of serious accidents in complex industrial systems such as at Three Mile Island and Bhopal has prompted development of new models of causation and investigation of disasters. These analytical models have potential relevance in anaesthesia. We therefore applied one of the previously described systems to the investigation of an anaesthetic accident. The model chosen describes two kinds of failures, both of which must be sought. The first group, active failures, consists of mistakes made by practitioners in the provision of care. The second group, latent failures, represents flaws in the administrative and productive system. The model emphasizes the search for latent failures and shows that prevention of active failures alone is insufficient to avoid further accidents if latent failures persist unchanged. These key features and the utility of this model are illustrated by application to a case of aspiration of gastric contents. While four active failures were recognized, an equal number of latent failures also became apparent. The identification of both types of failures permitted the formulation of recommendations to avoid further occurrences. Thus this model of accident causation can provide a useful mechanism to investigate and possibly prevent anaesthetic accidents. PMID:1544192

  19. Analysis of 121 fatal passenger car-adult pedestrian accidents in China.

    PubMed

    Zhao, Hui; Yin, Zhiyong; Yang, Guangyu; Che, Xingping; Xie, Jingru; Huang, Wei; Wang, Zhengguo

    2014-10-01

    To study the characteristics of fatal vehicle-pedestrian accidents in China,a team was established and passenger car-pedestrian crash cases occurring between 2006 and 2011 in Beijing and Chongqing, China were collected. A total of 121 fatal passenger car-adult pedestrian collisions were sampled and analyzed. The pedestrian injuries were scored according to Abbreviated Injury Scale (AIS) and Injury Severity Score (ISS). The demographical distributions of fatal pedestrian accidents differed from other pedestrian accidents. Among the victims, no significant discrepancy in the distribution of ISS and AIS in head, thorax, abdomen, and extremities by pedestrian age was found, while pedestrian behaviors prior to the crashes may affect the ISS. The distributions of AIS in head, thorax, and abdomen among the fatalities did not show any association with impact speeds or vehicle types, whereas there was a strong relationship between the ISS and impact speeds. Whether pedestrians died in the accident field or not was not associated with the ISS or AIS. The present results may be useful for not only forensic experts but also vehicle safety researchers. More investigations regarding fatal pedestrian accidents need be conducted in great detail. PMID:25287805

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

    SciTech Connect

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

    1996-12-01

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

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

    SciTech Connect

    Walston, S; Rowland, M; Campbell, K

    2011-07-27

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

  2. Timing analysis of PWR fuel pin failures

    SciTech Connect

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. ); Straka, M. )

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report.

  3. APT Blanket System Loss-of-Flow Accident (LOFA) Analysis Based on Initial Conceptual Design - Case 1: with Beam Shutdown and Active RHR

    SciTech Connect

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  4. Analysis of hospitalization occurred due to motorcycles accidents in São Paulo city

    PubMed Central

    Gorios, Carlos; Armond, Jane de Eston; Rodrigues, Cintia Leci; Pernambuco, Henrique; Iporre, Ramiro Ortiz; Colombo-Souza, Patrícia

    2015-01-01

    OBJECTIVE: To characterize the motorcycle accidents occurred in the city of São Paulo, SP, Brazil in the year 2013, with emphasis on information about hospital admissions from SIH/SUS. METHODS: This is a retrospective cross-sectional study. The study covered 5,597 motorcyclists traumatized in traffic accident during the year 2013 occurred in the city of São Paulo. A survey was conducted using secondary data from the Information System of Hospitalization Health System (SIH/SUS). RESULTS: In 2013, in the city of São Paulo there were 5,597 admissions of motorcyclists traumatized in traffic accidents, of which 89.8% were male. The admission diagnosis were: leg fracture, femur fracture, and intracranial injury. CONCLUSION: This study confirms other preliminary studies on several points, among which stands out the highest prevalence of male young adults. Level of Evidence II, Retrospective Study. PMID:26327804

  5. A Comprehensive Analysis of the X-15 Flight 3-65 Accident

    NASA Technical Reports Server (NTRS)

    Dennehy, Cornelius J.; Orr, Jeb S.; Barshi, Immanuel; Statler, Irving C.

    2014-01-01

    The November 15, 1967, loss of X-15 Flight 3-65-97 (hereafter referred to as Flight 3-65) was a unique incident in that it was the first and only aerospace flight accident involving loss of crew on a vehicle with an adaptive flight control system (AFCS). In addition, Flight 3-65 remains the only incidence of a single-pilot departure from controlled flight of a manned entry vehicle in a hypersonic flight regime. To mitigate risk to emerging aerospace systems, the NASA Engineering and Safety Center (NESC) proposed a comprehensive review of this accident. The goal of the assessment was to resolve lingering questions regarding the failure modes of the aircraft systems (including the AFCS) and thoroughly analyze the interactions among the human agents and autonomous systems that contributed to the loss of the pilot and aircraft. This document contains the outcome of the accident review.

  6. COBRA/TRAC analysis of the PKL reflood test K9

    SciTech Connect

    Wilkins, C.A.

    1982-03-01

    Experiments simulating the refill and reflood procedure after a loss-of-coolant accident (LOCA) in a 1300 MW PWR primary coolant system were performed at the KWU test facility in Erlangen, Germany. COBRA/TRAC was used to model experiment K9 of the PKL series (completed December 1979). PKL-K9 was characterized by a double-ended guillotine break in the cold leg with emergency core cooling water injected into the cold legs. COBRA/TRAC was able to successfully predict lower core temperature profiles and quench times, upper core temperature profiles until quench, upper plenum and break pressures and correct trends in collapsed water levels.

  7. APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break

    SciTech Connect

    Shadday, M.A.

    1999-11-19

    The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks.

  8. Comparative analysis of social, demographic, and flight-related attributes between accident and nonaccident general aviation pilots.

    PubMed

    Urban, R F

    1984-04-01

    This investigation represents an exploratory examination of several differentiating social and demographic characteristics for a sample of calendar year 1978 Colorado-resident nonfatal accident-involved pilots and a random sample of nonaccident general aviation (i.e., nonairline) pilots. During 1979-1980 80 currently active pilots were interviewed by the author, and information concerning the standard demographic variables, in addition to several social, psychological, and flying-related items, was obtained. The sample was generated from commercially available data files derived from U.S. Government records and consisted of 46 accident and 34 nonaccident pilots who resided within a 100-mi radius of Denver, east of the Rocky Mountains. Descriptively, the respondents represented a broad spectrum of general aviation, including: corporate pilots, "crop dusters," builders of amateur experimental aircraft, and recreational fliers. Application of stepwise discriminant analysis revealed that the pilots' education, political orientation, birth order, percent of flying for business purposes, participation in nonflying aviation activities, number of years of flying experience, and an index of aviation procedural noncompliance yielded statistically significant results. Furthermore, utilization of the classification capability of discriminant analysis produced a mathematical function which correctly allocated 78.5% of the cases into the appropriate groups, thus contributing to a 56.5% proportionate reduction in error over a random effects model. No relationship was found between accident involvement and several indicators of social attachments, socioeconomic status, and a number of measures of flying exposure. PMID:6732683

  9. Environmental risk management for radiological accidents: integrating risk assessment and decision analysis for remediation at different spatial scales.

    PubMed

    Yatsalo, Boris; Sullivan, Terrence; Didenko, Vladimir; Linkov, Igor

    2011-07-01

    The consequences of the Tohuku earthquake and subsequent tsunami in March 2011 caused a loss of power at the Fukushima Daiichi nuclear power plant, in Japan, and led to the release of radioactive materials into the environment. Although the full extent of the contamination is not currently known, the highly complex nature of the environmental contamination (radionuclides in water, soil, and agricultural produce) typical of nuclear accidents requires a detailed geospatial analysis of information with the ability to extrapolate across different scales with applications to risk assessment models and decision making support. This article briefly summarizes the approach used to inform risk-based land management and remediation decision making after the Chernobyl, Soviet Ukraine, accident in 1986. PMID:21608109

  10. Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

    SciTech Connect

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system {open_quotes}IMPACT{close_quotes} for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT`s distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data.

  11. Development of liquid-film tracking models for analysis of AP-600 passive containment cooling system

    SciTech Connect

    Sun, J.G.; Sha, W.T.; Chen, Y.S.

    1993-10-01

    The AP-600, an advanced pressurized water reactor, utilizes a passive containment cooling system (PCCS) to remove heat released inside the containment vessel following postulated design-base accidents (DBAS) such as a main-steam-line break or loss-of-coolant accident. The containment vessel consists of a vertical cylindrical shell and is capped at both top and bottom by a dome that is also a body of revolution with a meridian cross section in the shape of a semiellipse with a horizontal major axis. During a DBA, heat released to the interior of the steel containment vessel is removed by evaporation of a continuously flowing thin liquid film on the outside surface of the vessel, thus lowering the temperature of the steel vessel wall so that steam condenses on its inside surface. The external liquid film is formed by flooding water at top of the ellipsoidal dome. Evaporation of the falling liquid film is enhanced by buoyancy-driven flows of moist air in an annular space outside the steel containment vessel. To ensure PCCS performance, it is necessary to predict both the evaporating film on the outside surface of the vessel and the condensate film on its inside. To this end, two liquid-film tracking models for time-dependent flows (a simplified model and a comprehensive model) have been developed and implemented in the COMMIX code. COMMIX is a general-purpose, time-dependent, multidimensional computer code for thermal hydraulic analysis of single- or multicomponent engineering systems. It solves a system of conservation equations of continuities for up to six species, mixture momentum, and mixture energy, and a k-{epsilon} two-equation turbulent model. A unique feature of the COMMIX code is its porous-media formulation, which represents the first unified approach to thermal-hydraulic analysis. The tracking models discussed in this report compute the liquid-film thickness, its mean velocity, and its temperature on both sides of the steel containment vessel.

  12. Simplified containment event tree analysis for the Sequoyah Ice Condenser containment

    SciTech Connect

    Galyean, W.J.; Schroeder, J.A.; Pafford, D.J. )

    1990-12-01

    An evaluation of a Pressurized Water Reactor (PWR) ice condenser containment was performed. In this evaluation, simplified containment event trees (SCETs) were developed that utilized the vast storehouse of information generated by the NRC's Draft NUREG-1150 effort. Specifically, the computer programs and data files produced by the NUREG-1150 analysis of Sequoyah were used to electronically generate SCETs, as opposed to the NUREG-1150 accident progression event trees (APETs). This simplification was performed to allow graphic depiction of the SCETs in typical event tree format, which facilitates their understanding and use. SCETs were developed for five of the seven plant damage state groups (PDSGs) identified by the NUREG-1150 analyses, which includes: both short- and long-term station blackout sequences (SBOs), transients, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS). Steam generator tube rupture (SGTR) and event-V PDSGs were not analyzed because of their containment bypass nature. After being benchmarked with the APETs, in terms of containment failure mode and risk, the SCETs were used to evaluate a number of potential containment modifications. The modifications were examined for their potential to mitigate or prevent containment failure from hydrogen burns or direct impingement on the containment by the core, (both factors identified as significant contributors to risk in the NUREG-1150 Sequoyah analysis). However, because of the relatively low baseline risk postulated for Sequoyah (i.e., 12 person-rems per reactor year), none of the potential modifications appear to be cost effective. 15 refs., 10 figs. , 17 tabs.

  13. What can the drivers' own description from combined sources provide in an analysis of driver distraction and low vigilance in accident situations?

    PubMed

    Tivesten, Emma; Wiberg, Henrik

    2013-03-01

    Accident data play an important role in vehicle safety development. Accident data sources are generally limited in terms of how much information is provided on driver states and behaviour prior to an accident. However, the precise limitations vary between databases, due to differences in analysis focus and data collection procedures between organisations. If information about a specific accident can be retrieved from more than one data source it should be possible to combine the available information sets to facilitate data from one source to compensate for limitations in the other(s). To investigate the viability of such compensation, this study identified a set of accidents recorded in two different data sources. The first data source investigated was an accident mail survey and the second data source insurance claims documents consisting predominantly of insurance claims completed by the involved road users. An analysis of survey variables was compared to a case analysis including word data derived from the same survey and filed insurance claims documents. For each accident, the added value of having access to more than one source of information was assessed. To limit the scope of this study, three particular topics were investigated: available information on low vigilance (e.g., being drowsy, ill); secondary task distraction (e.g., talking with passengers, mobile phone use); and distraction related to the driving task (e.g., looking for approaching vehicles). Results suggest that for low vigilance and secondary task distraction, a combination of the mail survey and insurance claims documents provide more reliable and detailed pre-crash information than survey variables alone. However, driving related distraction appears to be more difficult to capture. In order to gain a better understanding of the above issues and how frequently they occur in accidents, the data sources and analysis methods suggested here may be combined with other investigation methods such

  14. Accident analysis and control options in support of the sludge water system safety analysis

    SciTech Connect

    HEY, B.E.

    2003-01-16

    A hazards analysis was initiated for the SWS in July 2001 (SNF-8626, K Basin Sludge and Water System Preliminary Hazard Analysis) and updated in December 2001 (SNF-10020 Rev. 0, Hazard Evaluation for KE Sludge and Water System - Project A16) based on conceptual design information for the Sludge Retrieval System (SRS) and 60% design information for the cask and container. SNF-10020 was again revised in September 2002 to incorporate new hazards identified from final design information and from a What-if/Checklist evaluation of operational steps. The process hazards, controls, and qualitative consequence and frequency estimates taken from these efforts have been incorporated into Revision 5 of HNF-3960, K Basins Hazards Analysis. The hazards identification process documented in the above referenced reports utilized standard industrial safety techniques (AIChE 1992, Guidelines for Hazard Evaluation Procedures) to systematically guide several interdisciplinary teams through the system using a pre-established set of process parameters (e.g., flow, temperature, pressure) and guide words (e.g., high, low, more, less). The teams generally included representation from the U.S. Department of Energy (DOE), K Basins Nuclear Safety, T Plant Nuclear Safety, K Basin Industrial Safety, fire protection, project engineering, operations, and facility engineering.

  15. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    SciTech Connect

    O`Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-12-31

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  16. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    SciTech Connect

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  17. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    SciTech Connect

    Dallman, R J; Gottula, R C; Holcomb, E E; Jouse, W C; Wagoner, S R; Wheatley, P D

    1987-05-01

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

  18. Launch Vehicle Fire Accident Preliminary Analysis of a Liquid-Metal Cooled Thermionic Nuclear Reactor: TOPAZ-II

    NASA Astrophysics Data System (ADS)

    Hu, G.; Zhao, S.; Ruan, K.

    2012-01-01

    In this paper, launch vehicle propellant fire accident analysis of TOPAZ-II reactor has been done by a thermionic reactor core analytic code-TATRHG(A) developed by author. When a rocket explodes on a launch pad, its payload-TOPAZ-II can be subjected to a severe thermal environment from the resulting fireball. The extreme temperatures associated with propellant fires can create a destructive environment in or near the fireball. Different kind of propellants - liquid propellant and solid propellant which will lead to different fire temperature are considered. Preliminary analysis shows that the solid propellant fires can melt the whole toxic beryllium radial reflector.

  19. Analysis of general aviation accidents during operations under instrument flight rules

    NASA Technical Reports Server (NTRS)

    Bennett, C. T.; Schwirzke, Martin; Harm, C.

    1990-01-01

    A report is presented to describe some of the errors that pilots make during flight under IFR. The data indicate that there is less risk during the approach and landing phase of IFR flights, as compared to VFR operations. Single-pilot IFR accident rates continue to be higher than two-pilot IFR incident rates, reflecting the high work load of IFR operations.

  20. Traffic accident in Cuiabá-MT: an analysis through the data mining technology.

    PubMed

    Galvão, Noemi Dreyer; de Fátima Marin, Heimar

    2010-01-01

    The traffic road accidents (ATT) are non-intentional events with an important magnitude worldwide, mainly in the urban centers. This article aims to analyzes data related to the victims of ATT recorded by the Justice Secretariat and Public Security (SEJUSP) in hospital morbidity and mortality incidence at the city of Cuiabá-MT during 2006, using data mining technology. An observational, retrospective and exploratory study of the secondary data bases was carried out. The three database selected were related using the probabilistic method, through the free software RecLink. One hundred and thirty-nine (139) real pairs of victims of ATT were obtained. In this related database the data mining technology was applied with the software WEKA using the Apriori algorithm. The result generated 10 best rules, six of them were considered according to the parameters established that indicated a useful and comprehensible knowledge to characterize the victims of accidents in Cuiabá. Finally, the findings of the associative rules showed peculiarities of the road traffic accident victims in Cuiabá and highlight the need of prevention measures in the collision accidents for males. PMID:20841739

  1. Spontaneous abortions after the Three Mile Island nuclear accident: a life table analysis.

    PubMed

    Goldhaber, M K; Staub, S L; Tokuhata, G K

    1983-07-01

    A study was conducted to determine whether the incidence of spontaneous abortion was greater than expected near the Three Mile Island (TMI) nuclear power plant during the months following the March 28, 1979 accident. All persons living within five miles of TMI were registered shortly after the accident, and information on pregnancy at the time of the accident was collected. After one year, all pregnancy cases were followed up and outcomes ascertained. Using the life table method, it was found that, given pregnancies after four completed weeks of gestation counting from the first day of the last menstrual period, the estimated incidence of spontaneous abortion (miscarriage before completion of 16 weeks of gestation) was 15.1 per cent for women pregnant at the time of the TMI accident. Combining spontaneous abortions and stillbirths (delivery of a dead fetus after 16 weeks of gestation), the estimated incidence was 16.1 per cent for pregnancies after four completed weeks of gestation. Both incidences are comparable to baseline studies of fetal loss. PMID:6859357

  2. Spontaneous abortions after the Three Mile Island nuclear accident: a life table analysis

    SciTech Connect

    Goldhaber, M.K.; Staub, S.L.; Tokuhata, G.K.

    1983-07-01

    A study was conducted to determine whether the incidence of spontaneous abortion was greater than expected near the Three Mile Island (TMI) nuclear power plant during the months following the March 28, 1979 accident. All persons living within five miles of TMI were registered shortly after the accident, and information on pregnancy at the time of the accident was collected. After one year, all pregnancy cases were followed up and outcomes ascertained. Using the life table method, it was found that, given pregnancies after four completed weeks of gestation counting from the first day of the last menstrual period, the estimated incidence of spontaneous abortion (miscarriage before completion of 16 weeks of gestation) was 15.1 per cent for women pregnant at the time of the TMI accident. Combining spontaneous abortions and stillbirths (delivery of a dead fetus after 16 weeks of gestation), the estimated incidence was 16.1 per cent for pregnancies after four completed weeks of gestation. Both incidences are comparable to baseline studies of fetal loss.

  3. 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.

  4. Risk Analysis for Public Consumption: Media Coverage of the Ginna Nuclear Reactor Accident.

    ERIC Educational Resources Information Center

    Dunwoody, Sharon; And Others

    Researchers have determined that the lay public makes risk judgments in ways that are very different from those advocated by scientists. Noting that these differences have caused considerable concern among those who promote and regulate health and safety, a study examined media coverage of the accident at the Robert E. Ginna nuclear power plant…

  5. Spontaneous abortions after the Three Mile Island nuclear accident: a life table analysis.

    PubMed Central

    Goldhaber, M K; Staub, S L; Tokuhata, G K

    1983-01-01

    A study was conducted to determine whether the incidence of spontaneous abortion was greater than expected near the Three Mile Island (TMI) nuclear power plant during the months following the March 28, 1979 accident. All persons living within five miles of TMI were registered shortly after the accident, and information on pregnancy at the time of the accident was collected. After one year, all pregnancy cases were followed up and outcomes ascertained. Using the life table method, it was found that, given pregnancies after four completed weeks of gestation counting from the first day of the last menstrual period, the estimated incidence of spontaneous abortion (miscarriage before completion of 16 weeks of gestation) was 15.1 per cent for women pregnant at the time of the TMI accident. Combining spontaneous abortions and stillbirths (delivery of a dead fetus after 16 weeks of gestation), the estimated incidence was 16.1 per cent for pregnancies after four completed weeks of gestation. Both incidences are comparable to baseline studies of fetal loss. PMID:6859357

  6. A Longitudinal Analysis of the Causal Factors in Major Maritime Accidents in the USA and Canada (1996-2006)

    NASA Technical Reports Server (NTRS)

    Johnson, C. W.; Holloway, C, M.

    2007-01-01

    Accident reports provide important insights into the causes and contributory factors leading to particular adverse events. In contrast, this paper provides an analysis that extends across the findings presented over ten years investigations into maritime accidents by both the US National Transportation Safety Board (NTSB) and Canadian Transportation Safety Board (TSB). The purpose of the study was to assess the comparative frequency of a range of causal factors in the reporting of adverse events. In order to communicate our findings, we introduce J-H graphs as a means of representing the proportion of causes and contributory factors associated with human error, equipment failure and other high level classifications in longitudinal studies of accident reports. Our results suggest the proportion of causal and contributory factors attributable to direct human error may be very much smaller than has been suggested elsewhere in the human factors literature. In contrast, more attention should be paid to wider systemic issues, including the managerial and regulatory context of maritime operations.

  7. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    SciTech Connect

    Gauntt, Randall O.; Mattie, Patrick D.

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  8. Retrospective reconstruction of Ioidne-131 distribution at the Fukushima Daiichi Nuclear Power Plant accident by analysis of Ioidne-129

    NASA Astrophysics Data System (ADS)

    Matsuzaki, Hiroyuki; Muramatsu, Yasuyuki; Toyama, Chiaki; Ohno, Takeshi; Kusuno, Haruka; Miyake, Yasuto; Honda, Maki

    2014-05-01

    Science and Education on June, 2011. So far more than 500 samples were measured and determined I-129 deposition amount by AMS at MALT (Micro Analysis Laboratory, Tandem accelerator), The University of Tokyo. The measurement error from AMS is less than 5%, typically 3%. The overall uncertainty is estimated less than 30%, including the uncertainty from that of the nominal value of the standard reference material used, that of I-129/I-131 ratio estimation, that of the "representativeness" for the region by the analyzed sample, etc. The isotopic ratio I-129/I-131 from the reactor was estimated [3] (to be 22.3 +- 6.3 as of March 11, 2011) from a series of samples collected by a group of The University of Tokyo on the 20th of April, 2011 for which the I-131 was determined by gamma-ray spectrometry with good precision. Complementarily, we had investigated the depth profile in soil of the accident derived I-129 and migration speed after the deposition and found that more than 90% of I-129 was concentrated within top 5 cm layer and the downward migration speed was less than 1cm/yr [4]. From the set of I-129 data, corresponding I-131 were calculated and the distribution map is going to be constructed. Various fine structures of the distribution came in sight. [1] Y. Nikiforov and D. R. Gnepp, 1994, Cancer, Vol. 47, pp748-766. [2] T. Straume, et al., 1996, Health Physics, Vol. 71, pp733-740. [3] Y. Miyake, H. Matsuzaki et al.,2012, Geochem. J., Vol. 46, pp327-333. [4] M. Honda, H. Matsuzaki et al., under submission.

  9. Thermal analysis of an irradiated-fuel concrete integrated container under normal and fire-accident conditions. Report No. 89-242-K

    SciTech Connect

    Taralis, D.

    1990-01-01

    This study describes the development of the special purpose three-dimensional heat transfer computer code for the thermal analysis of a Concrete Integrated Container (CIC) for the transportation of 10-year cooled fuel under normal conditions and hypothetical fire accident conditions. Results are given for: Comparisons of theoretical predictions with existing half-scale CIC experimental results, and representative analytical results for full-scale CIC under normal and fire accident conditions.

  10. 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.

  11. 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.

  12. 3D analysis of the reactivity insertion accident in VVER-1000

    SciTech Connect

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M.

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  13. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    SciTech Connect

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  14. 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.

  15. 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.

  16. An analysis of thermionic space nuclear reactor power system: I. Effect of disassembling radial reflector, following a reactivity initiated accident

    SciTech Connect

    El-Genk, M.S.; Paramonov, D. )

    1993-01-10

    An analysis is performed to determine the effect of disassembling the radial reflector of the TOPAZ-II reactor, following a hypothetical severe Reactivity Initiated Accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the drums to rotate the full 180[degree] outward at their maximum speed of 1.4[degree]/s. Results indicate that disassembling only three of twelve radial reflector panels would successfully shutdown the reactor, with little overheating of the fuel and the moderator.

  17. Analysis of Japanese Radionuclide Monitoring Data of Food Before and After the Fukushima Nuclear Accident

    PubMed Central

    2015-01-01

    In an unprecedented food monitoring campaign for radionuclides, the Japanese government took action to secure food safety after the Fukushima nuclear accident (Mar. 11, 2011). In this work we analyze a part of the immense data set, in particular radiocesium contaminations in food from the first year after the accident. Activity concentrations in vegetables peaked immediately after the campaign had commenced, but they decreased quickly, so that by early summer 2011 only a few samples exceeded the regulatory limits. Later, accumulating mushrooms and dried produce led to several exceedances of the limits again. Monitoring of meat started with significant delay, especially outside Fukushima prefecture. After a buildup period, contamination levels of meat peaked by July 2011 (beef). Levels then decreased quickly, but peaked again in September 2011, which was primarily due to boar meat (a known accumulator of radiocesium). Tap water was less contaminated; any restrictions for tap water were canceled by April 1, 2011. Pre-Fukushima 137Cs and 90Sr levels (resulting from atmospheric nuclear explosions) in food were typically lower than 0.5 Bq/kg, whereby meat was typically higher in 137Cs and vegetarian produce was usually higher in 90Sr. The correlation of background radiostrontium and radiocesium indicated that the regulatory assumption after the Fukushima accident of a maximum activity of 90Sr being 10% of the respective 137Cs concentrations may soon be at risk, as the 90Sr/137Cs ratio increases with time. This should be taken into account for the current Japanese food policy as the current regulation will soon underestimate the 90Sr content of Japanese foods. PMID:25621976

  18. SACO-1: a fast-running LMFBR accident-analysis code

    SciTech Connect

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  19. Analysis of reactivity-insertion accidents in the TREAT Upgrade reactor

    SciTech Connect

    Rudolph, R.R.; Bhattacharyya, S.K.

    1983-01-01

    The expansion of the experimental capabilities of the TREAT Upgrade (TU) reactor also tends to increase the potential risks associated with off-normal reactivity insertion incidents compared to the TREAT reactor. To provide adequate prtection for the public and the facility, while meeting experimenter's requirements, a specialized Reactor Trip System (RTS) with energy-dependent scram trips on reactor power and period has been developed. With this protection strategy, the consequences of reactivity insertion accidents in the TU reactor have been analyzed using a general methodology developed earlier. Results of these analyses are presented.

  20. Resolve! Version 2.5: Flammable Gas Accident Analysis Tool Acceptance Test Plan and Test Results

    SciTech Connect

    LAVENDER, J.C.

    2000-10-17

    RESOLVE! Version 2 .5 is designed to quantify the risk and uncertainty of combustion accidents in double-shell tanks (DSTs) and single-shell tanks (SSTs). The purpose of the acceptance testing is to ensure that all of the options and features of the computer code run; to verify that the calculated results are consistent with each other; and to evaluate the effects of the changes to the parameter values on the frequency and consequence trends associated with flammable gas deflagrations or detonations.

  1. Analysis of injuries among pilots involved in fatal general aviation airplane accidents.

    PubMed

    Wiegmann, Douglas A; Taneja, Narinder

    2003-07-01

    The purpose of this study was to analyze patterns of injuries sustained by pilots involved in fatal general aviation (GA) airplane accidents. Detailed information on the pattern and nature of injuries was retrieved from the Federal Aviation Administration's autopsy database for pilots involved in fatal GA airplane accidents from 1996 to 1999. A review of 559 autopsies revealed that blunt trauma was the primary cause of death in 86.0% (N=481) of the autopsies. The most commonly occurring bony injuries were fracture of the ribs (72.3%), skull (55.1%), facial bones (49.4%), tibia (37.9%) and pelvis (36.0%). Common organ injuries included laceration of the liver (48.1%), lung (37.6%) heart (35.6%), and spleen (30.1%), and hemorrhage of the brain (33.3%) and lung (32.9%). A fractured larynx was observed in 14.7% of the cases, a finding that has not been reported in literature until now. It was observed that individuals who sustained brain hemorrhage were also more likely to have fractures of the facial bones rather than skull fractures. PMID:12729820

  2. Thermal analysis of the 10-gallon and the 55-gallon DOT-6M containers with thermal boundary conditions corresponding to 10CFR71 normal transport and accident conditions

    SciTech Connect

    Sanchez, L.C.; Longenbaugh, R.S.; Moss, M.; Haseman, G.M.; Fowler, W.E.; Roth, E.P.

    1988-03-01

    This report describes the heat transfer analysis of the 10-gallon and 55-gallon 6M containers. The analysis was performed with boundary conditions corresponding to a normal transport condition and a hypothetical accident condition. Computational results indicated that the insulation material in the 6M containers will adequately protect the payload region of the 6M containers. 26 refs., 26 figs., 8 tabs.

  3. Health effects models for nuclear power plant accident consequence analysis. Part 1, Introduction, integration, and summary: Revision 2

    SciTech Connect

    Evans, J.S.; Abrahmson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.; Gilbert, E.S.

    1993-10-01

    This report is a revision of NUREG/CR-4214, Rev. 1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. This revision has been made to incorporate changes to the Health Effects Models recommended in two addenda to the NUREG/CR-4214, Rev. 1, Part 11, 1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR, ICRP and NAS/NRC (BEIR V). The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes are considered. Linear and linear-quadratic models are recommended for estimating the risks of seven types of cancer in adults - leukemia, bone, lung, breast, gastrointestinal, thyroid, and ``other``. For most cancers, both incidence and mortality are addressed. Five classes of genetic diseases -- dominant, x-linked, aneuploidy, unbalanced translocations, and multifactorial diseases are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk.

  4. 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.

  5. Analysis of containment venting for the Peach Bottom Atomic Power Station

    SciTech Connect

    Hanson, D.J.; Wright, R.E.; Jenkins, J.P.

    1986-09-12

    The effectiveness of containment venting as a means of preventing or mitigating the consequences of severe accidents was evaluated for Peach Bottom Atomic Power Station Units 2 and 3 (BWR-4s with Mark I containments). Results from this evaluation indicate that the effectiveness of venting in preventing containment failure is highly dependent on the severe accident sequence. Containment venting can be effective for several classes of sequences, including loss-of-coolant accidents with breaks in the containment and transients with a failure of containment heat removal. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with several sequences currently identified as significant contributors to risk. Means of improving the potential for risk reduction were identified, but their influence on risk was not analyzed.

  6. TRAC (Transient Reactor Analysis Code) model of reactor vent paths

    SciTech Connect

    Pevey, R.E.; Reece, J.W.

    1987-12-18

    The Safety Methods group of Scientific Computations Division (SCD) is currently calculating assembly power limits based on reactor response to a double-ended guillotine pipe break loss of coolant accident (LOCA). SCD has implemented a two-level approach in which the Transient Reactor Analysis Code (TRAC) is used to calculate the system pressure response to the LOCA, and these pressures serve as the boundary conditions for a detailed assembly calculation using FLOWTRAN. As part of the TRAC calculation, a detailed TRAC model of the reactor vent paths has been developed that involves the hardware in the top portion of the reactor tank through which air flows as the moderator tank drains following the LOCA initiation. The hardware included in this model are the top shield (with its many penetrations), the gas space above the top shield, the vacuum breakers, the U tube, the helium blanket gas system, and the gas ports. This detailed model is necessary for an accurate calculation of the tank pressures in the first few seconds of the LOCA because the initial tank depressurization is relieved through these vent paths. The tank pressures for about 5 seconds into the transient are sensitive to water flow from the gas space through the top shield, the associated expansion pressure drop of the blanket gas, and the clearing of the vacuum breakers and gas ports. This model was added to a previously developed TRAC model of the rest of the system and the resulting full system model was used to calculate the pressure response during the first few seconds of the LOCA. 8 refs., 8 figs.

  7. The Gulf of Mexico oil rig accident: analysis by different SAR satellite images

    NASA Astrophysics Data System (ADS)

    Del Frate, Fabio; Giacomini, Andrea; Latini, Daniele; Solimini, Domenico; Emery, William J.

    2011-11-01

    The management of the monitoring oil spills over the sea surface is a very important and actual task for international environmental agencies, due to the continuous risks represented by possible accidents involving either rigs or tankers. On the other hand the increase of remote sensing space missions can definitely improve our capabilities in this kind of activity. In this paper we consider the dramatic Gulf of Mexico oil spill event of 2010 to investigate on the types of information that could be provided by the available SAR images collection which included different polarizations and bands. With an eye to the implementation of fully automatic processing chains, an assessment of a novel segmentation technique based on PCNN (Pulse Coupled Neural Networks) was also carried out.

  8. Radiological health effects models for nuclear power plant accident consequence analysis.

    PubMed

    Evans, J S; Moeller, D W

    1989-04-01

    Improved health effects models have been developed for assessing the early effects, late somatic effects and genetic effects that might result from low-LET radiation exposures to populations following a major accident in a nuclear power plant. All the models have been developed in such a way that the dynamics of population risks can be analyzed. Estimates of life years lost and the duration of illnesses were generated and a framework recommended for summarizing health impacts. Uncertainty is addressed by providing models for upper, central and lower estimates of most effects. The models are believed to be a significant improvement over the models used in the U.S. Nuclear Regulatory Commission's Reactor Safety Study, and they can easily be modified to reflect advances in scientific understanding of the health effects of ionizing radiation. PMID:2925380

  9. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    SciTech Connect

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

  10. TRANSIENT ACCIDENT ANALYSIS OF THE GLOVEBOX SYSTEM IN A LARGE PROCESS ROOM

    SciTech Connect

    Lee, S

    2008-01-11

    Local transient hydrogen concentrations were evaluated inside a large process room when the hydrogen gas was released by three postulated accident scenarios associated with the process tank leakage and fire leading to a loss of gas confinement. The three cases considered in this work were fire in a room, loss of confinement from a process tank, and loss of confinement coupled with fire event. Based on these accident scenarios in a large and unventilated process room, the modeling calculations of the hydrogen migration were performed to estimate local transient concentrations of hydrogen due to the sudden leakage and release from a glovebox system associated with the process tank. The modeling domain represented the major features of the process room including the principal release or leakage source of gas storage system. The model was benchmarked against the literature results for key phenomena such as natural convection, turbulent behavior, gas mixing due to jet entrainment, and radiation cooling because these phenomena are closely related to the gas driving mechanisms within a large air space of the process room. The modeling results showed that at the corner of the process room, the gas concentrations migrated by the Case 2 and Case 3 scenarios reached the set-point value of high activity alarm in about 13 seconds, while the Case 1 scenario takes about 90 seconds to reach the concentration. The modeling results were used to estimate transient radioactive gas migrations in an enclosed process room installed with high activity alarm monitor when the postulated leakage scenarios are initiated without room ventilation.

  11. Atmospheric transport of radioactive debris to Norway in case of a hypothetical accident related to the recovery of the Russian submarine K-27.

    PubMed

    Bartnicki, Jerzy; Amundsen, Ingar; Brown, Justin; Hosseini, Ali; Hov, Øystein; Haakenstad, Hilde; Klein, Heiko; Lind, Ole Christian; Salbu, Brit; Szacinski Wendel, Cato C; Ytre-Eide, Martin Album

    2016-01-01

    The Russian nuclear submarine K-27 suffered a loss of coolant accident in 1968 and with nuclear fuel in both reactors it was scuttled in 1981 in the outer part of Stepovogo Bay located on the eastern coast of Novaya Zemlya. The inventory of spent nuclear fuel on board the submarine is of concern because it represents a potential source of radioactive contamination of the Kara Sea and a criticality accident with potential for long-range atmospheric transport of radioactive particles cannot be ruled out. To address these concerns and to provide a better basis for evaluating possible radiological impacts of potential releases in case a salvage operation is initiated, we assessed the atmospheric transport of radionuclides and deposition in Norway from a hypothetical criticality accident on board the K-27. To achieve this, a long term (33 years) meteorological database has been prepared and used for selection of the worst case meteorological scenarios for each of three selected locations of the potential accident. Next, the dispersion model SNAP was run with the source term for the worst-case accident scenario and selected meteorological scenarios. The results showed predictions to be very sensitive to the estimation of the source term for the worst-case accident and especially to the sizes and densities of released radioactive particles. The results indicated that a large area of Norway could be affected, but that the deposition in Northern Norway would be considerably higher than in other areas of the country. The simulations showed that deposition from the worst-case scenario of a hypothetical K-27 accident would be at least two orders of magnitude lower than the deposition observed in Norway following the Chernobyl accident. PMID:25804322

  12. An aftermath analysis of the 2014 coal mine accident in Soma, Turkey: Use of risk performance indicators based on historical experience.

    PubMed

    Spada, Matteo; Burgherr, Peter

    2016-02-01

    On the 13th of May 2014 a fire related incident in the Soma coal mine in Turkey caused 301 fatalities and more than 80 injuries. This has been the largest coal mine accident in Turkey, and in the OECD country group, so far. This study investigated if such a disastrous event should be expected, in a statistical sense, based on historical observations. For this purpose, PSI's ENSAD database is used to extract accident data for the period 1970-2014. Four different cases are analyzed, i.e., OECD, OECD w/o Turkey, Turkey and USA. Analysis of temporal trends for annual numbers of accidents and fatalities indicated a non-significant decreasing tendency for OECD and OECD w/o Turkey and a significant one for USA, whereas for Turkey both measures showed an increase over time. The expectation analysis revealed clearly that an event with the consequences of the Soma accident is rather unlikely for OECD, OECD w/o Turkey and USA. In contrast, such a severe accident has a substantially higher expectation for Turkey, i.e. it cannot be considered an extremely rare event, based on historical experience. This indicates a need for improved safety measures and stricter regulations in the Turkish coal mining sector in order to get closer to the rest of OECD. PMID:26687539

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

    SciTech Connect

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

    1994-01-01

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

  14. Fault tree analysis of fire and explosion accidents for dual fuel (diesel/natural gas) ship engine rooms

    NASA Astrophysics Data System (ADS)

    Guan, Yifeng; Zhao, Jie; Shi, Tengfei; Zhu, Peipei

    2016-07-01

    In recent years, China's increased interest in environmental protection has led to a promotion of energy-efficient dual fuel (diesel/natural gas) ships in Chinese inland rivers. A natural gas as ship fuel may pose dangers of fire and explosion if a gas leak occurs. If explosions or fires occur in the engine rooms of a ship, heavy damage and losses will be incurred. In this paper, a fault tree model is presented that considers both fires and explosions in a dual fuel ship; in this model, dual fuel engine rooms are the top events. All the basic events along with the minimum cut sets are obtained through the analysis. The primary factors that affect accidents involving fires and explosions are determined by calculating the degree of structure importance of the basic events. According to these results, corresponding measures are proposed to ensure and improve the safety and reliability of Chinese inland dual fuel ships.

  15. Tools for improving safety management in the Norwegian Fishing Fleet occupational accidents analysis period of 1998-2006.

    PubMed

    Aasjord, Halvard L

    2006-01-01

    Reporting of human accidents in the Norwegian Fishing Fleet has always been very difficult because there has been no tradition in making reports on all types of working accidents among fishermen, if the accident does not seem to be very serious or there is no economical incentive to report. Therefore reports are only written when the accidents are serious or if the fisherman is reported sick. Reports about an accident are sent to the insurance company, but another report should also be sent to the Norwegian Maritime Directorate (NMD). Comparing of data from one former insurance company and NMD shows that the real numbers of injuries or serious accidents among Norwegian fishermen could be up to two times more than the numbers reported to NMD. Special analyses of 1690 accidents from the so called PUS-database (NMD) for the period 1998-2002, show that the calculated risk was 23.6 accidents per 1000 man-years. This is quite a high risk level, and most of the accidents in the fishing fleet were rather serious. The calculated risks are highest for fishermen on board the deep sea fleet of trawlers (28.6 accidents per 1000 man-years) and also on the deep sea fleet of purse seiners (28.9 accidents per 1000 man-years). Fatal accidents over a longer period of 51.5 years from 1955 to 2006 are also roughly analysed. These data from SINTEF's own database show that the numbers of fatal accidents have been decreasing over this long period, except for the two periods 1980-84 and 1990-94 where we had some casualties with total losses of larger vessels with the loss of most of the crew, but also many others typical work accidents on smaller vessels. The total numbers of registered Norwegian fishermen and also the numbers of man-years have been drastically reduced over the 51.5 years from 1955 to 2006. The risks of fatal accidents have been very steady over time at a high level, although there has been a marked risk reduction since 1990-94. For the last 8.5-year period of January 1998

  16. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    SciTech Connect

    Harper, F.T.; Young, M.L.; Miller, L.A.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes.

  17. Episode analysis of deposition of radiocesium from the Fukushima Daiichi nuclear power plant accident.

    PubMed

    Morino, Yu; Ohara, Toshimasa; Watanabe, Mirai; Hayashi, Seiji; Nishizawa, Masato

    2013-03-01

    Chemical transport models played key roles in understanding the atmospheric behaviors and deposition patterns of radioactive materials emitted from the Fukushima Daiichi nuclear power plant after the nuclear accident that accompanied the great Tohoku earthquake and tsunami on 11 March 2011. However, model results could not be sufficiently evaluated because of limited observational data. We assess the model performance to simulate the deposition patterns of radiocesium ((137)Cs) by making use of airborne monitoring survey data for the first time. We conducted ten sensitivity simulations to evaluate the atmospheric model uncertainties associated with key model settings including emission data and wet deposition modules. We found that simulation using emissions estimated with a regional-scale (∼ 500 km) model better reproduced the observed (137)Cs deposition pattern in eastern Japan than simulation using emissions estimated with local-scale (∼ 50 km) or global-scale models. In addition, simulation using a process-based wet deposition module reproduced the observations well, whereas simulation using scavenging coefficients showed large uncertainties associated with empirical parameters. The best-available simulation reproduced the observed (137)Cs deposition rates in high-deposition areas (≥ 10 kBq m(-2)) within 1 order of magnitude and showed that deposition of radiocesium over land occurred predominantly during 15-16, 20-23, and 30-31 March 2011. PMID:23391028

  18. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

    SciTech Connect

    Evans, J.S. . School of Public Health)

    1990-01-01

    This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, models are included for assessing the risks of several nonlethal early and continuing effects -- including prodromal vomiting and diarrhea, hypothyroidism and radiation thyroiditis, skin burns, reproductive effects, and pregnancy losses. Linear and linear-quadratic models are recommended for estimating cancer risks. Parameters are given for analyzing the risks of seven types of cancer in adults -- leukemia, bone, lung, breast, gastrointestinal, thyroid, and other.'' The category, other'' cancers, is intended to reflect the combined risks of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus and cervix. Models of childhood cancers due to in utero exposure are also developed. For most cancers, both incidence and mortality are addressed. The models of cancer risk are derived largely from information summarized in BEIR III -- with some adjustment to reflect more recent studies. 64 refs., 18 figs., 46 tabs.

  19. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    SciTech Connect

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  20. One- and two-dimensional STEALTH simulations of fuel-pin transient response. Final report. [BWR; PWR

    SciTech Connect

    Wahi, K.K.

    1980-08-01

    This report presents an assessment of the adaptability of EPRI's one- and two-dimensional STEALTH computer codes to perform transient fuel rod analysis. The ability of the STEALTH code to simulate transient mechanical or thermomechanical loss-of-coolant accident is described. Analytic models of one- and two-dimensional formulations and features included in the two-dimensional simulation are discussed.

  1. Spatiotemporal Analysis for Wildlife-Vehicle Based on Accident Statistics of the County Straubing-Bogen in Lower Bavaria

    NASA Astrophysics Data System (ADS)

    Pagany, R.; Dorner, W.

    2016-06-01

    During the last years the numbers of wildlife-vehicle-collisions (WVC) in Bavaria increased considerably. Despite the statistical registration of WVC and preventive measures at areas of risk along the roads, the number of such accidents could not be contained. Using geospatial analysis on WVC data of the last five years for county Straubing-Bogen, Bavaria, a small-scale methodology was found to analyse the risk of WVC along the roads in the investigated area. Various indicators were examined, which may be related to WVC. The risk depends on the time of the day and year which shows correlations in turn to the traffic density and wildlife population. Additionally the location of the collision depends on the species and on different environmental parameters. Accidents seem to correlate with the land use left and right of the street. Land use data and current vegetation were derived from remote sensing data, providing information of the general land use, also considering the vegetation period. For this a number of hot spots was selected to identify potential dependencies between land use, vegetation and season. First results from these hotspots show, that WVCs do not only depend on land use, but may show a correlation with the vegetation period. With regard to agriculture and seasonal as well as annual changes this indicates that warnings will fail due to their static character in contrast to the dynamic situation of land use and resulting risk for WVCs. This shows that there is a demand for remote sensing data with a high spatial and temporal resolution as well as a methodology to derive WVC warnings considering land use and vegetation. With remote sensing data, it could become possible to classify land use and calculate risk levels for WVC. Additional parameters, derived from remote sensed data that could be considered are relief and crops as well as other parameters such as ponds, natural and infrastructural barriers that could be related to animal behaviour and

  2. [Accidents and acts of violence in Brazil: I--Analysis of mortality data].

    PubMed

    Jorge, M H; Gawryszewski, V P; Latorre, M do R

    1997-08-01

    External causes are an important cause of death in almost all countries. They are always the second or third in the mortality ranking, but their distribution according to type varies from country to country. Mortality due to external causes by type, gender and age, for Brazil as a whole and for state capitals specifically, is analysed. Mortality rates and the proportional mortality from 1977 to 1994 were calculated. The results showed that the number of deaths due to external causes has almost doubled from 1977 to 1994 and nowadays this is the second cause of death in Brazil. The mortality rate, in 1991, was 69.8 per 100,000 inhabitants and the highest increase was in the male rates. The male rates are almost 4.5 times greater than the female ones. The first cause of death among people from 5 to 39 years old is external causes, and the majority occur between 15 and 19 years of age (65% of the deaths by external causes). Besides the growth in itself it also seems that a shift of deaths to hower ages is occurring. Both mortality by traffic accidents and that by homicide have increased over the period from 1977 to 1994. Suicides have been stable and "other external causes" have increased slowly, especially due to falls and drowning. The mortality rates for external causes in state capitals are higher than the average for Brazil as a whole, except for some northeastern capitals. The rates for the capitals in the northern region are the highest in Brazil. In the northeastern region, only Recife, Maceió and Salvador have high rates. In the southeast, Vitória, Rio de Janeiro and S. Paulo have the highest rates in the country but Belo Horizonte's rates are declining. In the southern region all the capitals showed a growth in the rates as also in the capitals of the West-central region. The growth of mortality due to external causes type of external cause is different in these capitals. Suicide is not a public health problem in Brazil nor the state capitals. Traffic

  3. Modeling and sensitivity analysis of transport and deposition of radionuclides from the Fukushima Daiichi accident

    NASA Astrophysics Data System (ADS)

    Hu, X.; Li, D.; Huang, H.; Shen, S.; Bou-Zeid, E.

    2014-01-01

    The atmospheric transport and ground deposition of radioactive isotopes 131I and 137Cs during and after the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident (March 2011) are investigated using the Weather Research and Forecasting/Chemistry (WRF/Chem) model. The aim is to assess the skill of WRF in simulating these processes and the sensitivity of the model's performance to various parameterizations of unresolved physics. The WRF/Chem model is first upgraded by implementing a radioactive decay term into the advection-diffusion solver and adding three parameterizations for dry deposition and two parameterizations for wet deposition. Different microphysics and horizontal turbulent diffusion schemes are then tested for their ability to reproduce observed meteorological conditions. Subsequently, the influence on the simulated transport and deposition of the characteristics of the emission source, including the emission rate, the gas partitioning of 131I and the size distribution of 137Cs, is examined. The results show that the model can predict the wind fields and rainfall realistically. The ground deposition of the radionuclides can also potentially be captured well but it is very sensitive to the emission characterization. It is found that the total deposition is most influenced by the emission rate for both 131I and 137Cs; while it is less sensitive to the dry deposition parameterizations. Moreover, for 131I, the deposition is also sensitive to the microphysics schemes, the horizontal diffusion schemes, gas partitioning and wet deposition parameterizations; while for 137Cs, the deposition is very sensitive to the microphysics schemes and wet deposition parameterizations, and it is also sensitive to the horizontal diffusion schemes and the size distribution.

  4. ADAPT (Analysis of Dynamic Accident Progression Trees) Beta Version 0.9

    Energy Science and Technology Software Center (ESTSC)

    2010-01-07

    The purpose of the ADAPT code is to generate Dynamic Event Trees (DET) using a user specified simulator. ADAPT can utilize any simulation tool which meets a minimal set of requirements. ADAPT is based on the concept of DET which use explicit modeling of the deterministic dynamic processes that take place during a nuclear reactor plant system evolution along with stochastic modeling. When DET are used to model different aspects of Probabilistic Risk Assessment (PRA),more » all accident progression scenarios starting from an initiating event are considered simultaneously. The DET branching occurs at user specified times and/or when an action is required by the system and/or the operator. These outcomes then decide how the dynamic system variables will evolve in time for each DET branch. Since two different outcomes at a DET branching may lead to completely different paths for system evolution, the next branching for these paths may occur not only at different times, but can be based on different branching criteria. The computational infrastructure allows for flexibility in ADAPT to link with different system simulation codes, parallel processing of the scenarios under consideration, on-line scenario management (initiation as well as termination) and user friendly graphical capabilities. The ADAPT system is designed for a distributed computing environment; the scheduler can track multiple concurrent branches simultaneously. The scheduler is modularized so that the DET branching strategy can be modified (e.g. biasing towards the worse case scenario/event). Independent database systems store data from the simulation tasks and the DET structure so that the event tree can be constructed and analyzed later. ADAPT is provided with a user-friendly client which can easily sort through and display the results of an experiment, precluding the need for the user to manually inspect individual simulator runs.« less

  5. Material distribution in light water reactor-type bundles tested under severe accident conditions

    SciTech Connect

    Noack, V.; Hagen, S.J.L.; Hofmann, P.; Schanz, G.; Sepold, L.K.

    1997-02-01

    Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorber and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.

  6. 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.

  7. 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.

  8. Accident analysis for transuranic waste management alternatives in the U.S. Department of Energy waste management program

    SciTech Connect

    Nabelssi, B.; Mueller, C.; Roglans-Ribas, J.; Folga, S.; Tompkins, M.; Jackson, R.

    1995-03-01

    Preliminary accident analyses and radiological source term evaluations have been conducted for transuranic waste (TRUW) as part of the US Department of Energy (DOE) effort to manage storage, treatment, and disposal of radioactive wastes at its various sites. The approach to assessing radiological releases from facility accidents was developed in support of the Office of Environmental Management Programmatic Environmental Impact Statement (EM PEIS). The methodology developed in this work is in accordance with the latest DOE guidelines, which consider the spectrum of possible accident scenarios in the implementation of various actions evaluated in an EIS. The radiological releases from potential risk-dominant accidents in storage and treatment facilities considered in the EM PEIS TRUW alternatives are described in this paper. The results show that significant releases can be predicted for only the most severe and extremely improbable accidents sequences.

  9. Supplemental analysis of accident sequences and source terms for waste treatment and storage operations and related facilities for the US Department of Energy waste management programmatic environmental impact statement

    SciTech Connect

    Folga, S.; Mueller, C.; Nabelssi, B.; Kohout, E.; Mishima, J.

    1996-12-01

    This report presents supplemental information for the document Analysis of Accident Sequences and Source Terms at Waste Treatment, Storage, and Disposal Facilities for Waste Generated by US Department of Energy Waste Management Operations. Additional technical support information is supplied concerning treatment of transuranic waste by incineration and considering the Alternative Organic Treatment option for low-level mixed waste. The latest respirable airborne release fraction values published by the US Department of Energy for use in accident analysis have been used and are included as Appendix D, where respirable airborne release fraction is defined as the fraction of material exposed to accident stresses that could become airborne as a result of the accident. A set of dominant waste treatment processes and accident scenarios was selected for a screening-process analysis. A subset of results (release source terms) from this analysis is presented.

  10. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    SciTech Connect

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  11. Shipping container response to severe highway and railway accident conditions: Appendices

    SciTech Connect

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  12. 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

  13. [Safety assessment of a Brazilian company based on analysis of work accidents by the causal tree method].

    PubMed

    Binder, M C; Pham, D; de Almeida, I M

    1998-01-01

    We present here the results of a study of 21 work-related accidents that occurred in a Brazilian manufacturing company. The aim was to assess the safety level of the company to improve its work accident prevention policy. In the last 6 months of 1992 and 1993, all accidents resulting in 15 days' absence from work, reported for social security purposes, were analyzed using the INRS causal tree method (ADC) and a questionnaire completed on site. Potential risk factors for accidents were identified based on the specific factors highlighted by the ADC. More universal trees were also compiled for the safety assessment. Three hundred and thirty specific accident factors were recorded (mean of 15.71 per accident). This is consistent with there being multiple causes of accidents rather than the assertion of Brazilian business safety departments that accidents are due to "dangerous" or "unsafe" behavior. Introducing the idea of culpability into accidents prevents the implementation of an appropriate information feedback process, essential for effective prevention. However, the large number of accidents related to "material" (78%) and "environment" (70%) indicates that working conditions are poor. This shows that the technical risks, mostly due to unsafe machinery and equipment are not being dealt with. Seventy-five potential accident factors were identified. Of these, 35% were "organizational", a high proportion for the company studied. Improvisation occurs at all levels, particularly at the organizational level. This is thus a major determinant for entire series of, if not most, accident situations. The poor condition of equipment also plays a major role in accidents. The effects of poor equipment on safety exacerbate the organizational shortcomings. The company's safety intervention policy should improve the management of human resources (rules designating particular workers for particular workstations; instructions for the safe operation of machines and equipment

  14. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    SciTech Connect

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

  15. Fire accident analysis modeling in support of non-reactor nuclear facilities at Sandia National Laboratories

    SciTech Connect

    Restrepo, L.F.

    1993-06-01

    The Department of Energy (DOE) requires that fire hazard analyses (FHAs) be conducted for all nuclear and new facilities, with results for the latter incorporated into Title I design. For those facilities requiring safety analysis documentation, the FHA shall be documented in the Safety Analysis Reports (SARs). This paper provides an overview of the methodologies and codes being used to support FHAs at Sandia facilities, with emphasis on SARs.

  16. 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

  17. 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.

  18. 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…

  19. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

  20. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  1. Preliminary Accident Analysis for Construction and Operation of the Chornobyl New Safety Confinement

    SciTech Connect

    Batiy, Valeriy; Rubezhansky, Yruiy; Rudko, Vladimir; shcherbin, vladimir; Yegorov, V; Schmieman, Eric A.; Timmins, Douglas C.

    2005-08-08

    Analysis of potential exposure of personal and population during construction and exploitation of the New Safe Confinement was made. Scenarios of hazard event development were ranked. It is shown, that as a whole construction and exploitation of the NSC are in accordance with actual radiation safety norms of Ukraine.

  2. Temporal uncertainty analysis of human errors based on interrelationships among multiple factors: a case of Minuteman III missile accident.

    PubMed

    Rong, Hao; Tian, Jin; Zhao, Tingdi

    2016-01-01

    In traditional approaches of human reliability assessment (HRA), the definition of the error producing conditions (EPCs) and the supporting guidance are such that some of the conditions (especially organizational or managerial conditions) can hardly be included, and thus the analysis is burdened with incomprehensiveness without reflecting the temporal trend of human reliability. A method based on system dynamics (SD), which highlights interrelationships among technical and organizational aspects that may contribute to human errors, is presented to facilitate quantitatively estimating the human error probability (HEP) and its related variables changing over time in a long period. Taking the Minuteman III missile accident in 2008 as a case, the proposed HRA method is applied to assess HEP during missile operations over 50 years by analyzing the interactions among the variables involved in human-related risks; also the critical factors are determined in terms of impact that the variables have on risks in different time periods. It is indicated that both technical and organizational aspects should be focused on to minimize human errors in a long run. PMID:26360211

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

    SciTech Connect

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

    1990-06-01

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

  4. 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)

  5. 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.

  6. A working man`s analysis of incidents and accidents with explosives at the Los Alamos National Laboratory, 1946--1997

    SciTech Connect

    Ramsay, J.B.; Goldie, R.H.

    1998-12-31

    At the inception of the Laboratory hectic and intense work was the norm during the development of the atomic bombs. After the war the development of other weapons for the Cold War again contributed to an intense work environment. Formal Standard Operating Procedures (SOPs) were not required at that time. However, the occurrence of six fatalities in 1959 during the development of a new high-energy plastic bonded explosive (94% HMX) forced the introduction SOPs. After an accident at the Department of Energy (DOE) plant at Amarillo, TX in 1977, the DOE promulgated the Department wide DOE Explosives Safety Manual. Table 1 outlines the history of the introduction of SOPs and the DOE Explosives Safety Manual. Many of the rules and guidelines presented in these documents were developed and introduced as the result of an incident or accident. However, many of the current staff are not familiar with the background of the development. To preserve as much of this knowledge as possible, they are collecting documentation on incidents and accidents involving energetic materials at Los Alamos. Formal investigations of serious accidents elucidate the multiple causes that contributed to accidents. These reports are generally buried in a file and, and are not read by more recent workers. Reports involving fatalities at Los Alamos before 1974 were withheld from the general employee. Also, these documents contain much detail and analysis that is not of interest to the field worker. The authors have collected the documents describing 116 incidents and have analyzed the contributing factors as viewed from the standpoint of the individual operator. All the incidents occurred at the Los Alamos National Laboratory and involved energetic materials in some manner, though not all occurred within the explosive handling groups. Most accidents are caused by multiple contributing factors. They have attempted to select the one or two factors that they consider as the most important relative to the

  7. Analysis of avalanche risk factors in backcountry terrain based on usage frequency and accident data in Switzerland

    NASA Astrophysics Data System (ADS)

    Techel, F.; Zweifel, B.; Winkler, K.

    2015-09-01

    Recreational activities in snow-covered mountainous terrain in the backcountry account for the vast majority of avalanche accidents. Studies analyzing avalanche risk mostly rely on accident statistics without considering exposure (or the elements at risk), i.e., how many, when and where people are recreating, as data on recreational activity in the winter mountains are scarce. To fill this gap, we explored volunteered geographic information on two social media mountaineering websites - bergportal.ch and camptocamp.org. Based on these data, we present a spatiotemporal pattern of winter backcountry touring activity in the Swiss Alps and compare this with accident statistics. Geographically, activity was concentrated in Alpine regions relatively close to the main Swiss population centers in the west and north. In contrast, accidents occurred equally often in the less-frequented inner-alpine regions. Weekends, weather and avalanche conditions influenced the number of recreationists, while the odds to be involved in a severe avalanche accident did not depend on weekends or weather conditions. However, the likelihood of being involved in an accident increased with increasing avalanche danger level, but also with a more unfavorable snowpack containing persistent weak layers (also referred to as an old snow problem). In fact, the most critical situation for backcountry recreationists and professionals occurred on days and in regions when both the avalanche danger was critical and when the snowpack contained persistent weak layers. The frequently occurring geographical pattern of a more unfavorable snowpack structure also explains the relatively high proportion of accidents in the less-frequented inner-alpine regions. These results have practical implications: avalanche forecasters should clearly communicate the avalanche danger and the avalanche problem to the backcountry user, particularly if persistent weak layers are of concern. Professionals and recreationists, on the

  8. A probabilistic analysis of a catastrophic transuranic waste hoist accident at the WIPP

    SciTech Connect

    Greenfield, M.A. |; Sargent, T.J. |

    1993-06-01

    This report builds upon the extensive and careful analyses made by the DOE of the probability of failure of the waste hoist, and more particularly on the probability of failure of a major component, the hydraulic brake system. The extensive fault tree analysis prepared by the DOE was the starting point of the present report. A key element of this work is the use of probability distributions rather than so-called point estimates to describe the probability of failure of an element. One of the authors (MAG) developed the expressions for the probability of failure of the brake system. The second author (TJS) executed the calculations of the final expressions for failure probabilities. The authors hope that this work will be of use to the DOE in its evaluation of the safety of the waste hoist, a key element at the WIPP.

  9. Factors Associated with Fatal Occupational Accidents among Mexican Workers: A National Analysis

    PubMed Central

    Gonzalez-Delgado, Mery; Gómez-Dantés, Héctor; Fernández-Niño, Julián Alfredo; Robles, Eduardo; Borja, Víctor H.; Aguilar, Miriam

    2015-01-01

    Objective To identify the factors associated with fatal occupational injuries in Mexico in 2012 among workers affiliated with the Mexican Social Security Institute. Methods Analysis of secondary data using information from the National Occupational Risk Information System, with the consequence of the occupational injury (fatal versus non-fatal) as the response variable. The analysis included 406,222 non-fatal and 1,140 fatal injuries from 2012. The factors associated with the lethality of the injury were identified using a logistic regression model with the Firth approach. Results Being male (OR=5.86; CI95%: 4.22-8.14), age (OR=1.04; CI95%: 1.03-1.06), employed in the position for 1 to 10 years (versus less than 1 year) (OR=1.37; CI95%: 1.15-1.63), working as a facilities or machine operator or assembler (OR: 3.28; CI95%: 2.12- 5.07) and being a worker without qualifications (OR=1.96; CI95%: 1.18-3.24) (versus an office worker) were associated with fatality in the event of an injury. Additionally, companies classified as maximum risk (OR=1.90; CI 95%: 1.38-2.62), workplace conditions (OR=7.15; CI95%: 3.63-14.10) and factors related to the work environment (OR=9.18; CI95%:4.36-19.33) were identified as risk factors for fatality in the event of an occupational injury. Conclusions Fatality in the event of an occupational injury is associated with factors related to sociodemographics (age, sex and occupation), the work environment and workplace conditions. Worker protection policies should be created for groups with a higher risk of fatal occupational injuries in Mexico. PMID:25790063

  10. Emulation and Sobol' sensitivity analysis of an atmospheric dispersion model applied to the Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Girard, Sylvain; Mallet, Vivien; Korsakissok, Irène; Mathieu, Anne

    2016-04-01

    Simulations of the atmospheric dispersion of radionuclides involve large uncertainties originating from the limited knowledge of meteorological input data, composition, amount and timing of emissions, and some model parameters. The estimation of these uncertainties is an essential complement to modeling for decision making in case of an accidental release. We have studied the relative influence of a set of uncertain inputs on several outputs from the Eulerian model Polyphemus/Polair3D on the Fukushima case. We chose to use the variance-based sensitivity analysis method of Sobol'. This method requires a large number of model evaluations which was not achievable directly due to the high computational cost of Polyphemus/Polair3D. To circumvent this issue, we built a mathematical approximation of the model using Gaussian process emulation. We observed that aggregated outputs are mainly driven by the amount of emitted radionuclides, while local outputs are mostly sensitive to wind perturbations. The release height is notably influential, but only in the vicinity of the source. Finally, averaging either spatially or temporally tends to cancel out interactions between uncertain inputs.

  11. 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

  12. 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

  13. 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.

  14. ARIES-ACT1 Safety Design and Analysis

    SciTech Connect

    Humrickhouse, Paul W.; Merrill, Brad J.

    2014-01-01

    ARIES-ACT1 (Advanced and Conservative Tokamak) is a 1000-MW(electric) tokamak design featuring advanced plasma physics and divertor and blanket engineering. Some relevant features include an advanced SiC blanket with PbLi as coolant and breeder; a helium-cooled steel structural ring and tungsten divertors; a thin-walled, helium-cooled vacuum vessel; and a room-temperature, water-cooled shield outside the vacuum vessel. We consider here some safety aspects of the ARIES-ACT1 design and model a series of design-basis and beyond-design-basis accidents with the MELCOR code modified for fusion. The presence of multiple coolants (PbLi, helium, and water) makes possible a variety of such accidents. We consider here a loss-of-flow accident caused by a long-term station blackout (LTSBO), an ex-vessel helium break into the cryostat, and a beyond-design-basis accident in which a LTSBO is aggravated by a loss-of-coolant accident in ARIES-ACT1's ultimate decay heat removal system, the water-cooled shield. In the design-basis accidents, we find that the secondary confinement boundaries are not challenged, and the structural integrity of in-vessel components is not threatened by high temperatures or pressures; decay heat can be passively removed.

  15. 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.

  16. 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

  17. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    Energy Science and Technology Software Center (ESTSC)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized watermore » reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.« less

  18. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    SciTech Connect

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized water reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.

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

    SciTech Connect

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

    1984-04-01

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

  20. Analysis of Road Traffic Accident Rate in the Slovak Republic and Possibilities of Its Reduction through Telematic Applications

    NASA Astrophysics Data System (ADS)

    Kalašová, Alica; Krchová, Zuzana

    There is a worldwide trend of increasing number of vehicles being newly introduced on an annual basis. On the road, there are great numbers of drivers with low discipline, aggressive driving habits and non-compliance with basic principles of responsible traffic behavior. As a consequence, the number of traffic accidents is rising, and so do their repercussions.

  1. Code System for Fluid-Structure Interaction Analysis.

    Energy Science and Technology Software Center (ESTSC)

    2001-05-30

    Version 00 PELE-IC is a two-dimensional semi-implicit Eulerian hydrodynamics program for the solution of incompressible flow coupled to flexible structures. The code was developed to calculate fluid-structure interactions and bubble dynamics of a pressure-suppression system following a loss-of-coolant accident (LOCA). The fluid, structure, and coupling algorithms have been verified by calculation of benchmark problems and air and steam blowdown experiments. The code is written for both plane and cylindrical coordinates. The coupling algorithm is generalmore » enough to handle a wide variety of structural shapes. The concepts of void fractions and interface orientation are used to track the movement of free surfaces, allowing great versatility in following fluid-gas interfaces both for bubble definition and water surface motion without the use of marker particles.« less

  2. A real-time, dynamic early-warning model based on uncertainty analysis and risk assessment for sudden water pollution accidents.

    PubMed

    Hou, Dibo; Ge, Xiaofan; Huang, Pingjie; Zhang, Guangxin; Loáiciga, Hugo

    2014-01-01

    A real-time, dynamic, early-warning model (EP-risk model) is proposed to cope with sudden water quality pollution accidents affecting downstream areas with raw-water intakes (denoted as EPs). The EP-risk model outputs the risk level of water pollution at the EP by calculating the likelihood of pollution and evaluating the impact of pollution. A generalized form of the EP-risk model for river pollution accidents based on Monte Carlo simulation, the analytic hierarchy process (AHP) method, and the risk matrix method is proposed. The likelihood of water pollution at the EP is calculated by the Monte Carlo method, which is used for uncertainty analysis of pollutants' transport in rivers. The impact of water pollution at the EP is evaluated by expert knowledge and the results of Monte Carlo simulation based on the analytic hierarchy process. The final risk level of water pollution at the EP is determined by the risk matrix method. A case study of the proposed method is illustrated with a phenol spill accident in China. PMID:24781332

  3. Methodology and computational framework used for the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement accident analysis

    SciTech Connect

    Mueller, C.; Roglans-Ribas, J.; Folga, S.; Huttenga, A.; Jackson, R.; TenBrook, W.; Russell, J. |

    1994-02-01

    A methodology, computational framework, and integrated PC-based database have been developed to assess the risks of facility accidents in support of the US Department of Energy (DOE) Environmental Restoration and Waste Management Programmatic Environmental Impact Statement. The methodology includes the following interrelated elements: (1) screening of storage and treatment processes and related waste inventories to determine risk-dominant facilities across the DOE complex, (2) development and frequency estimation of the risk-dominant sequences of accidents, and (3) determination of the evolution of and final compositions of radiological or chemically hazardous source terms predicted to be released as a function of the storage inventory or treatment process throughput. The computational framework automates these elements to provide source term input for the second part of the analysis which includes (1) development or integration of existing site-specific demographics and meteorological data and calculation of attendant unit-risk factors and (2) assessment of the radiological or toxicological consequences of accident releases to the general public and to the occupational work force.

  4. Analysis of 129I in the soils of Fukushima Prefecture: preliminary reconstruction of 131I deposition related to the accident at Fukushima Daiichi Nuclear Power Plant (FDNPP).

    PubMed

    Muramatsu, Yasuyuki; Matsuzaki, Hiroyuki; Toyama, Chiaki; Ohno, Takeshi

    2015-01-01

    Iodine-131 is one of the most critical radionuclides to be monitored after release from reactor accidents due to the tendency for this nuclide to accumulate in the human thyroid gland. However, there are not enough data related to the reactor accident in Fukushima, Japan to provide regional information on the deposition of this short-lived nuclide (half-life = 8.02 d). In this study we have focused on the long-lived iodine isotope, (129)I (half-life of 1.57 × 10(7) y), and analyzed it by accelerator mass spectrometry (AMS) for surface soil samples collected at various locations in Fukushima Prefecture. In order to obtain information on the (131)I/(129)I ratio released from the accident, we have determined (129)I concentrations in 82 soil samples in which (131)I concentrations were previously determined. There was a strong correlation (R(2) = 0.84) between the two nuclides, suggesting that the (131)I levels in soil samples following the accident can be estimated through the analysis of (129)I. We have also examined the possible influence from (129m)Te on (129)I, and found no significant effect. In order to construct a deposition map of (131)I, we determined the (129)I concentrations (Bq/kg) in 388 soil samples collected from different locations in Fukushima Prefecture and the deposition densities (Bq/m(2)) of (131)I were reconstructed from the results. PMID:24930438

  5. Quantification and uncertainty analysis of source terms for severe accidents in light water reactors (QUASAR): Part 2, Sensitivity analysis techniques

    SciTech Connect

    Ishigami, T.; Cazzoli, E.; Khatib-Rahbar, M.; Unwin, S.D.

    1987-10-01

    Existing methods for sensitivity analysis are described and new techniques are proposed. These techniques are evaluated through consideration relative to the QUASAR program. Merits and limitations of the various approaches are examined by a detailed application to the Suppression Pool Aerosol Removal Code (SPARC). 17 refs., 7 figs., 12 tabs.

  6. Assessment of ISLOCA risk-methodology and application to a combustion engineering plant

    SciTech Connect

    Kelly, D.L.; Auflick, J.L.; Haney, L.N.

    1992-04-01

    Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISOLOCA core damage frequency and risk. This report presents a detailed of description of the application of this analysis methodology to a Combustion Engineering plant.

  7. Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Bury, Tomasz

    2011-12-01

    Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

  8. 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

  9. 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

  10. 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)

  11. Analysis of the reasons for accidents and of protective measures against induced voltage on aerial electrical transmission lines

    SciTech Connect

    Misrikhanov, M. Sh.; Mirzaabdullaev, A. O.

    2009-01-15

    The problem of safety during work on aerial transmission lines under an induced voltage is examined. Results are presented from a study of the causes of accidents over the last 20 years in electrical grids in this country. A determination of different levels of induced voltage on disconnected aerial transmission lines as a function of their grounding scheme is proposed. The order of magnitudes for each level are given, along with approximate expressions for calculating them.

  12. Calculation notes that support accident scenario and consequence development for the steam intrusion from interfacing systems accident

    SciTech Connect

    Ryan, G.W., Westinghouse Hanford

    1996-07-25

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Steam Intrusion from Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included within.

  13. A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete Interaction

    SciTech Connect

    Farmer, M. T.

    2015-11-01

    The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO2.

  14. 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.

  15. 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.

  16. 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.

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

    SciTech Connect

    Rebak, Raul B.

    2014-12-30

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

  18. 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)

  19. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    SciTech Connect

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

  20. 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.