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Sample records for accident loca simulation

  1. 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.

  2. 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.

  3. 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.

  4. Effect of aging on EPR cable electrical performance during LOCA simulations. [Ethylene propylene rubber

    SciTech Connect

    Bustard, L.D.

    1984-01-01

    When exposed to a LOCA environment, some ethylene propylene rubber (EPR) cable materials experience substantial moisture absorption and dimensional changes. These phenomena may contribute to mechanical damage of the cable insulation resulting in electrical degradation. Recent experiments illustrate that the extent of moisture absorption and dimensional changes during an accident simulation are dependent on the EPR product, the accelerated age, and the aging technique employed to achieve that age. Results for several commercial EPR materials are summarized.

  5. Effect of LOCA simulation procedures on cross-linked polyolefin cable's performance

    SciTech Connect

    Bustard, L.D.

    1984-04-01

    Electrical and mechanical properties of three commercial cross-linked polyolefin (XLPO) materials, typically used as electrical cable insulation, have been monitored during these simulations of nuclear power plant aging and accident stresses. For one XLPO cable accelerated thermal aging is performed, then the samples are irradiated to the combined aging and LOCA total dose. Finally, a steam exposure is applied. For a second and third set of XLPO cables simultaneous applications of elevated temperature and radiation stresses are used to preaccident age specimens. These aging exposures are followed by simultaneous and steam exposures to simulate a LOCA environment. The measurement parameters during these tests included: dc insulation resistance, ac leakage current, ultimate tensile strength, ultimate tensile elongation, percentage dimensional changes, and percentage moisture absorption. Test results for three XLPO materials are presented.

  6. Effect of LOCA simulation procedures on ethylene propylene rubber's mechanical and electrical properties

    SciTech Connect

    Bustard, L.D.

    1983-10-01

    Electrical and mechanical properties of several commercial ethylene-propylene rubber (EPR) materials, typically used as electrical cable insulation, have been monitored during three simulations of nuclear power plant aging and accident stresses. For one set of cables and separate tensile specimens we did a sequential test. We first performed accelerated thermal aging, then irradiated the samples to the combined aging and LOCA total dose. Finally we applied a steam exposure. For a second and third set of cables and separate tensile specimens we used simultaneous applications of elevated temperature and radiation stresses to preaccident age our specimens. We followed these aging exposures by simultaneous radiation and steam exposures to simulate a LOCA environment. Our measurement parameters during these tests included: dc insulation resistance, ac leakage current, ultimate tensile strength, ultimate tensile elongation, percentage dimensional changes, and percentage moisture absorption. We present test results for nine EPR materials. The implications of our research results for future cable qualification testing efforts is discussed.

  7. 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.

  8. Transient deformation properties of Zircaloy for LOCA simulation. Final report

    SciTech Connect

    Hann, C. R.; Mohr, C. L.; Busness, K. M.; Olson, N. J.; Reich, F. R.; Stewart, K. B.

    1980-05-01

    This experimental data report is Volume 4 of a series of 5 volumes describing the oxidation and deformation rate behavior of Zircaloy cladding under simulated LOCA conditions. It contains listings of strain versus stress, time, and temperature evaluated from the numerical constitutive relationships and the original data used to develop them. This volume also contains listings of the ramp load, pressure, and temperature test data from both current and previous phases of the series, as well as material describing applications of the data.

  9. Zircoloy Cladding Oxidation Simulation for LWR under LOCA Conditions

    2003-04-25

    PRECIP-2 simulates zircaloy cladding oxidation under LOCA conditions of LWR’s. The code calculates oxygen concentration distribution across the cladding wall by solving the diffusion equation with moving boundary conditions, taking into account the structure change of the beta— phase, i.e. alpha precipitation during the cooling period. The code also predicts total oxygen uptake, thicknesses of alpha, beta and oxide layers.

  10. 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.

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

    SciTech Connect

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

    2013-05-06

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

  12. 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. LOCA simulation in NRU program: data report for the fourth materials experiment (MT-4)

    SciTech Connect

    Wilson, C.L.; Mohr, C.L.; Hesson, G.M.; Wildung, N.J.; Russcher, G.E.; Webb, B.J.; Freshley, M.D.

    1983-07-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program by Pacific Northwest Laboratory (PNL). This experiment (MT-4) was funded by the US Nuclear Regulatory Commission (NRC) to evaluate ballooning and rupture during adiabatic heatup in the temperature range of 1033 to 1200K (1400 to 1700/sup 0/F). The 12 rest rods in the center of the 32-rod bundle were initially pressurized to 4.62 MPa (670 psia) to insure rupture in the correct temperature range. All 12 test rods ruptured with an average strain of 43.7% at the maximum flow blockage elevation of 2.68 m (105.4 in.). Experimental data for the MT-4 transient experiment and post-test measurements and photographs of the fuel are presented in this report.

  14. Materials Test-2 LOCA Simulation in the NRU Reactor

    SciTech Connect

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

  15. LOCA simulation in the NRU reactor: materials test-1

    SciTech Connect

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

  16. 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.

  17. 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.

  18. Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR

    SciTech Connect

    Chapman, R.H.; Crowley, J.L.; Longest, A.W.

    1982-01-01

    Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive to large deformation were selected in order to favor rod-to-rod interactions. The tests showed that rod-to-rod interactions play an important role in the deformation process.

  19. Severe accident simulation at Olkiuoto

    SciTech Connect

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

    1995-09-01

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

  20. 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

  1. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    SciTech Connect

    GrandJean, C.; Cauvin, R.; Lebuffe, C.

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  2. 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.

  3. 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.

  4. 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.

  5. SPACE code simulation of cold leg small break LOCA in the ATLAS integral test

    SciTech Connect

    Kim, B. J.; Kim, H. T.; Kim, J.; Kim, K. D.

    2012-07-01

    SPACE code is a system analysis code for pressurized water reactors. This code uses a two-fluid and three-field model. For a few years, intensive validations have been performed to secure the prediction accuracy of models and correlations for two-phase flow and heat transfer. Recently, the code version 1.0 was released. This study is to see how well SPACE code predicts thermal hydraulic phenomena of an integral effect test. The target experiment is a cold leg small break LOCA in the ATLAS facility, which has the same two-loop features as APR1400. Predicted parameters were compared with experimental observations. (authors)

  6. LOCA Simulation in the National Research Universal Reactor Program Postirradiation Examination Results for the Third Materials Experiment (MT-3) - Second Campaign

    SciTech Connect

    Haberman, J. H.

    1985-06-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR} fuel bundles as part of the Loss-of-Coolant Accident (LOCA} Simulation Program by Pacific Northwest Laboratory (PNL). The third materials test (MT-3} was the sixth experiment in a series of thermalhydraulic and materials deformation/rupture experiments conducted in the National Research Universal (NRU) Reactor, Chalk River, Ontario, Canada. The MT-3 experiment was jointly funded by the U.S. Nuclear Regulatory Commission (NRC) and the United Kingdom Atomic Energy Authority (UKAEA) with the main objective of evaluating ballooning and rupture during active two-phase cooling at elevated temperatures. All 12 test rods in the center of the 32-rod bundle failed with an average peak strain of 55.4%. At the request of the UKAEA, a destructive postirradiation examination (PIE) was performed on 7 of the 12 test rods. The results of this examination were presented in a previous report. Subsequently, and at the request of UKAEA, PIE was performed on three additional rods along with further examination of one of the previously examined rods. Information obtained from the PIE included cladding thickness measurements, cladding metallography, and particle size analysis of the fractured fuel pellets. This report describes the additional PIE work performed and presents the results of the examinations.

  7. Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

    SciTech Connect

    Wilson, C.L.; Hesson, G.M.; Pilger, J.P.; King, L.L.; Panisko, F.E.

    1993-09-01

    This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuel bundle is cooled.

  8. 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.

  9. Effect of alternative aging and accident simulations on polymer properties

    SciTech Connect

    Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

    1984-01-01

    The response of eighteen US and French polymer materials to variations in aging and accident simulation techniques has been determined by this experimental program. This information will provide a partial data base by which to judge appropriate simulation practices. The overall research goal was to determine whether some aging and accident simulation techniques are better suited for qualification activities than other alternative simulation techniques.

  10. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect

    El-Genk, Mohamed

    2012-10-19

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

  11. 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.

  12. Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

    SciTech Connect

    Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

    1981-07-01

    The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

  13. 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.

  14. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    SciTech Connect

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

    1998-03-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs.

  15. Effect of alternative aging and accident simulations on polymer properties

    SciTech Connect

    Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

    1985-05-01

    The influence of accident irradiation, steam, and chemical spray exposures on the behavior of twenty-three age-preconditioned polymer sample sets (twenty-one different materials) has been investigated. The test program varied the following conditions: (1) Accident simulations of irradiation and thermodynamic (steam and chemical spray) conditions were performed both sequentially and simultaneously. (2) Accident thermodynamic (steam and chemical spray) exposures were performed both with and without air present during the exposures. (3) Sequential accident irradiations were performed both at 28/sup 0/C and 70/sup 0/C. (4) Age preconditioning was performed both sequentially and simultaneously. (5) Sequential aging irradiations were performed both at 27/sup 0/C and 70/sup 0/C. (6) Sequential aging exposures were performed using two sequences: (1) thermal followed by irradiation and (2) irradiation followed by thermal. We report both general trends applicable to a majority of the tested materials as well as specific results for each polymer. Our data base consists of ultimate tensile properties at the completion of the accident exposure for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. We also report bend test results at completion of the accident exposures for two TEFZEL materials and permanent set after compression results for three EPR, one VAMAC, one BUNA N, one SILICONE, and one VITON material.

  16. Response of HEPA filters to simulated-accident conditions

    SciTech Connect

    Gregory, W.S.; Martin, R.A.; Smith, P.R.; Fenton, D.E.

    1982-01-01

    High-efficiency particulate air (HEPA) filters have been subjected to simulated accident conditions to determine their response to abnormal operating events. Both domestic and European standard and high-capacity filters have been evaluated to determine their response to simulated fire, explosion, and tornado conditions. The HEPA filter structural limitations for tornado and explosive loadings are discussed. In addition, filtration efficiencies during these accident conditions are reported for the first time. Our data indicate efficiencies between 80% and 90% for shock loadings below the structural limit level. We describe two types of testing for ineffective filtration - clean filters exposed to pulse-entrained aerosol and dirty filters exposed to tornado and shock pulses. Efficiency and material loss data are described. Also, the resonse of standard HEPA filters to simulated fire conditions is presented. We describe a unique method of measuring accumulated combustion products on the filter. Additionally, data relating to pressure drop vs accumulated mass during plugging are reported for simulated combustion aerosols. The effects of concentration and moisture levels on filter plugging were evaluated. We are obtaining all of the above data so that mathematical models can be developed for fire, explosion, and tornado accident analysis computer codes. These computer codes can be used to assess the response of nuclear air cleaning systems to accident conditions.

  17. Simulation of Accident Sequences Including Emergency Operating Procedures

    SciTech Connect

    Queral, Cesar; Exposito, Antonio; Hortal, Javier

    2004-07-01

    Operator actions play an important role in accident sequences. However, design analysis (Safety Analysis Report, SAR) seldom includes consideration of operator actions, although they are required by compulsory Emergency Operating Procedures (EOP) to perform some checks and actions from the very beginning of the accident. The basic aim of the project is to develop a procedure validation system which consists of the combination of three elements: a plant transient simulation code TRETA (a C based modular program) developed by the CSN, a computerized procedure system COPMA-III (Java technology based program) developed by the OECD-Halden Reactor Project and adapted for simulation with the contribution of our group and a software interface that provides the communication between COPMA-III and TRETA. The new combined system is going to be applied in a pilot study in order to analyze sequences initiated by secondary side breaks in a Pressurized Water Reactors (PWR) plant. (authors)

  18. 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.

  19. Assessment of Environmental Qualification Practices and Condition Monitoring Techniques for Low-Voltage Electric Cables: LOCA Test Results

    SciTech Connect

    Lofaro, R.; Grove, E.; Villaran, M.; Soo, P.; Hsu, F.

    2001-02-01

    This report documents the results of a research program addressing issues related to the qualification process for low-voltage instrumentation and control (I&C) electric cables used in commercial nuclear power plants. Three commonly used types of I&C cable were tested: Cross-Linked Polyethylene (XLPE) insulation with a Neoprene® jacket, Ethylene Propylene Rubber (EPR) insulation with an unbonded Hypalon® jacket, and EPR with a bonded Hypalon® jacket. Each cable type received accelerated aging to simulate 20, 40, and 60 years of qualified life. In addition, naturally aged cables of the same types were obtained from decommissioned nuclear power plants and tested. The cables were subjected to simulated loss-of-coolant-accident (LOCA) conditions, which included the sequential application of LOCA radiation followed by exposure to steam at high temperature and pressure, as well as to chemical spray. Periodic condition monitoring (CM) was performed using nine different techniques to obtain data on the condition of the cable, as well as to evaluate the effectiveness of those CM techniques for in situ monitoring of cables. Volume 1 of this report presents the results of the LOCA tests, and Volume 2 discusses the results of the condition monitoring tests.

  20. 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

  1. Skills development at a paramedic accident simulation centre.

    PubMed

    Donaghy, John

    2016-02-01

    Practice simulation in acute and pre-hospital care settings is a growing area of interest for clinicians and health educationalists, and there is much evidence to support its use (Pike and O'Donnell 2010). Most simulation is delivered through computer-aided software or in virtual environments, however last year the University of Hertfordshire opened an accident simulation centre which is an outdoor facility that offers pre- and post-registration paramedics the opportunity to experience a range of scenarios in a 'real life' but secure environment. This article describes how the centre enables students to apply theory to practice in complex situations, such as managing patients injured in road traffic collisions. PMID:26853672

  2. Skills development at a paramedic accident simulation centre.

    PubMed

    Donaghy, John

    2016-02-01

    Practice simulation in acute and pre-hospital care settings is a growing area of interest for clinicians and health educationalists, and there is much evidence to support its use (Pike and O'Donnell 2010). Most simulation is delivered through computer-aided software or in virtual environments, however last year the University of Hertfordshire opened an accident simulation centre which is an outdoor facility that offers pre- and post-registration paramedics the opportunity to experience a range of scenarios in a 'real life' but secure environment. This article describes how the centre enables students to apply theory to practice in complex situations, such as managing patients injured in road traffic collisions.

  3. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    SciTech Connect

    A. K. MAJI; B. MARSHALL; ET AL

    2000-10-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  4. Simulation Study of Traffic Accidents in Bidirectional Traffic Models

    NASA Astrophysics Data System (ADS)

    Moussa, Najem

    Conditions for the occurrence of bidirectional collisions are developed based on the Simon-Gutowitz bidirectional traffic model. Three types of dangerous situations can occur in this model. We analyze those corresponding to head-on collision; rear-end collision and lane-changing collision. Using Monte Carlo simulations, we compute the probability of the occurrence of these collisions for different values of the oncoming cars' density. It is found that the risk of collisions is important when the density of cars in one lane is small and that of the other lane is high enough. The influence of different proportions of heavy vehicles is also studied. We found that heavy vehicles cause an important reduction of traffic flow on the home lane and provoke an increase of the risk of car accidents.

  5. Ethylene propylene cable degradation during LOCA research tests: tensile properties at the completion of accelerated aging

    SciTech Connect

    Bustard, L.D.

    1982-05-01

    Six ethylene-propylene rubber (EPR) insulation materials were aged at elevated temperature and radiation stress exposures common in cable LOCA qualification tests. Material samples were subjected to various simultaneous and sequential aging simulations in preparation for accident environmental exposures. Tensile properties subsequent to the aging exposure sequences are reported. The tensile properties of some, but not all, specimens were sensitive to the order of radiation and elevated temperature stress exposure. Other specimens showed more severe degradation when simultaneously exposed to radiation and elevated temperature as opposed to the sequential exposure to the same stresses. Results illustrate the difficulty in defining a single test procedure for nuclear safety-related qualification of EPR elastomers. A common worst-case sequential aging sequence could not be identified.

  6. A catastrophe-theory model for simulating behavioral accidents

    SciTech Connect

    Souder, W.E.

    1988-01-01

    Behavioral accidents are a particular type of accident. They are caused by inappropriate individual behaviors and faulty reactions. Catastrophe theory is a means for mathematically modeling the dynamic processes that underlie behavioral accidents. Based on a comprehensive data base of mining accidents, a computerized catastrophe model has been developed by the Bureau of Mines. This model systematically links individual psychological, group behavioral, and mine environmental variables with other accident causing factors. It answers several longstanding questions about why some normally safe behaving persons may spontaneously engage in unsafe acts that have high risks of serious injury. Field tests with the model indicate that it has three imnportant uses: it can be used as a effective training aid for increasing employee safety consciousness; it can be used as a management laboratory for testing decision alternatives and policies; and it can be used to help design the most effective work teams.

  7. Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

    NASA Astrophysics Data System (ADS)

    Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

    2016-04-01

    Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

  8. System code requirements for SBWR LOCA predictions

    SciTech Connect

    Rohatgi, U.S.; Slovik, G.; Kroeger, P.

    1994-12-31

    The simplified boiling water reactor (SBWR) is the latest design in the family of boiling water reactors (BWRs) from General Electric. The concept is based on many innovative, passive, safety systems that rely on naturally occurring phenomena, such as natural circulation, gravity flows, and condensation. Reliability has been improved by eliminating active systems such as pumps and valves. The reactor pressure vessel (RPV) is connected to heat exchangers submerged in individual water tanks, which are open to atmosphere. These heat exchanger, or isolation condensers (ICs), provide a heat sink to reduce the RPV pressure when isolated. The RPV is also connected to three elevated tanks of water called the gravity-driven cooling system (GDCS). During a loss-of-coolant accident (LOCA), the RPV is depressurized by the automatic depressurization system (ADS), allowing the gravity-driven flow from the GDCS tanks. The containment pressure is controlled by a passive containment cooling system (PCCS) and suppression pool. Similarly, there are new plant protection systems in the SBWR, such as fine-motion control rod drive, passive standby liquid control system, and the automatic feedwater runback system. These safety and plant protection systems respond to phenomena that are different from previous BWR designs. System codes must be upgraded to include models for the phenomena expected during transients for the SBWR.

  9. Comparison of passive safety and the safety injection systems under loss of coolant accident

    NASA Astrophysics Data System (ADS)

    Tahir, M.; Chughtai, I. R.; Lodhi, M. A. K.

    2009-04-01

    A Passive Safety Injection System (PSIS) and a Safety Injection System (SIS) with reference to a typical pressurized water reactor have been studied. The performance of the PSIS has been analyzed for a large break Loss of Coolant Accident (LOCA) in one of the cold leg of reactor coolant system. The SIS is a huge system consisting of many active components needing electrical power to perform its role of core cooling as high head safety injection system under designed accidents. The PSIS consist of passive components and performs its function automatically under gravity. In a reactor transient simulation, the PSIS and the SIS are tested for large break LOCA under the same boundary conditions. Critical thermal hydraulic parameters of both the systems are presented. Results obtained are approximately similar in both cases. Nevertheless, the PSIS would be a better choice for handling such scenarios due to its reduced and passive components.

  10. A catastrophe-theory model for simulating behavioral accidents

    SciTech Connect

    Souder, W.E.

    1988-01-01

    Based on a comprehensive data base of mining accidents, a computerized catastrophe model has been developed by the Bureau of Mines. This model systematically links individual psychological, group behavioral, and mine environmental variables with other accident causing factors. It answers several longstanding questions about why some normally safe behaving persons may spontaneously engage in unsafe acts that have high risks of serious injury. Field tests with the model indicate that it has three important uses: It can be used as an effective training aid for increasing employee safety consciousness; it can be used as a management laboratory for testing decision alternatives and policies; and it can be used to help design the most effective work teams.

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

    SciTech Connect

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

    1985-01-01

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

  12. Overview of the M5{sup R} Alloy behavior under RIA and LOCA Conditions

    SciTech Connect

    Mardon, J.P.; Dunn, B.

    2007-07-01

    Experience from irradiation in PWRs has confirmed the M5{sup R} possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. In this paper accident behavior is demonstrated through a comparison of M5{sup R} and Zircaloy-4 cladding behavior under RIA (Reactivity Insertion Accident) and LOCA (Loss Of Coolant Accident) conditions. AREVA NP supports a significant experimental program of analytical and full -scale tests along with comprehensive analyses on both M5{sup R} and SRA low-tin Zircaloy-4. A key presumption in the conduct of such tests is that, for all Zirconium alloys, the primary effects of high burn-up on cladding thermal-mechanical properties arise from the accumulation of hydrogen within the cladding during operation. This hypothesis is supported through a summarisation of the results of the main RIA and LOCA tests performed on virgin, pre-hydrided, and irradiated M5{sup R} and SRA low-tin Zircaloy-4 cladding. The first part of the paper presents the results of recent Room Temperature (RT) and High Temperature High Pressure (HTHP) integral RIA tests, mainly from the NSRR and CABRI programs, and separate effects mechanical properties tests on high burn-up M5{sup R} and Zircaloy- 4 irradiated claddings. In the second part of this paper, studies of cladding performance under LOCA conditions are presented.. The discussion includes high temperature oxidation kinetics, quench behaviour and post quenched mechanical behaviour of virgin, pre-hydrided and irradiated M5{sup R} and Zircaloy-4 cladding tubes after oxidation at LOCA temperatures and various quenching scenarios. The hydrogen concentrations studied are alloy dependent. Included are mechanical tests and in-depth metallurgical investigations developed to understand the failure mechanisms with the differing alloys and hydrogen concentrations. The result is a confirmation that the effect of hydrogen uptake dominates on the RIA and LOCA

  13. Simulation of three lanes one-way freeway in low visibility weather by possible traffic accidents

    NASA Astrophysics Data System (ADS)

    Pang, Ming-bao; Zheng, Sha-sha; Cai, Zhang-hui

    2015-09-01

    The aim of this work is to investigate the traffic impact of low visibility weather on a freeway including the fraction of real vehicle rear-end accidents and road traffic capacity. Based on symmetric two-lane Nagel-Schreckenberg (STNS) model, a cellular automaton model of three-lane freeway mainline with the real occurrence of rear-end accidents in low visibility weather, which considers delayed reaction time and deceleration restriction, was established with access to real-time traffic information of intelligent transportation system (ITS). The characteristics of traffic flow in different visibility weather were discussed via the simulation experiments. The results indicate that incoming flow control (decreasing upstream traffic volume) and inputting variable speed limits (VSL) signal are effective in accident reducing and road actual traffic volume's enhancing. According to different visibility and traffic demand the appropriate control strategies should be adopted in order to not only decrease the probability of vehicle accidents but also avoid congestion.

  14. Simulation study of traffic car accidents at a single lane roundabout

    NASA Astrophysics Data System (ADS)

    Echab, H.; Lakouari, N.; Ez-Zahraouy, H.; Benyoussef, A.

    2016-07-01

    In this paper, using the Nagel-Schreckenberg model, we numerically investigate the probability Pac of entering/circulating car accidents to occur at single-lane roundabout under the expanded open boundary. The roundabout consists of N on-ramps (respectively, off-ramps). The boundary is controlled by the injecting rates α1,α2 and the extracting rate β. The simulation results show that, depending on the injecting rates, the car accidents are more likely to happen when the capacity of the rotary is set to its maximum. Moreover, we found that the large values of rotary size L and the probability of preferential Pexit are reliable to improve safety and reduce accidents. However, the usage of indicator, the increase of β and/or N provokes an increase of car accident probability.

  15. APRIL.MOD3X - An interactive computer simulator of severe accidents in BWRs

    SciTech Connect

    Podowski, M.Z.; Kurul, N.; Lahey, R.T. Jr.; Burger, J.M.

    1996-12-31

    APRIL is a fast-running and user-friendly system code for interactive simulations of severe accidents in boiling water reactors (BWRs). The component models in the most recent version, APRIL.MOD3X, include the reactor core and pressure vessel, as well as the primary and secondary containments. Whereas APRIL.MOD3X is a fast-running code, the models are mechanistically based and account for several important local phenomena affecting accident progression. The code has been extensively validated against experimental data. The code can be run either in a stand-alone fashion or in conjunction with its graphical user interface (GUI). APRIL.MOD3X is intended for use in developing improved accident management strategies in support of emergency preparedness procedures and in assessment of the consequences of postulated accident scenarios.

  16. 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.

  17. 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.

  18. Two-lane traffic simulations with a blockage induced by an accident car

    NASA Astrophysics Data System (ADS)

    Zhu, H. B.; Lei, L.; Dai, S. Q.

    2009-07-01

    Based on the two-lane traffic model proposed by Chowdhury et al., a highway traffic model with a blockage induced by an accident car is proposed, in which both symmetric lane changing rules and asymmetric lane changing rules are adopted. The fundamental diagrams and spatial-temporal profiles are presented after the numerical simulation and the jam transition is studied. It is shown that the accident car not only causes a local jam behind the accident car, but also causes vehicles to cluster in the bypass lane. The asymmetric lane changing rules are more advantageous in reducing the local jam than the symmetric lane changing rules when the accident car is in the right lane, and the symmetric lane changing rules are superior when the accident car is in the left lane. Furthermore the curves of lane-changing frequency against the total density are given. It is found that the vehicles will change lane more frequently when traffic is inhomogeneous with different types of vehicle or with an accident car.

  19. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    SciTech Connect

    Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

    1994-09-01

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER.

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

    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

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

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

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

  2. Calculation of fuel pin failure timing under LOCA conditions

    SciTech Connect

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

    1991-10-01

    The objective of this research was 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) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs.

  3. Simulation of atmospheric dispersion of radioactivity from the Chernobyl accident

    SciTech Connect

    Lange, R.; Sullivan, T.J.; Gudiksen, P.H. )

    1989-07-01

    Measurements of airborne radioactivity over Europe, Japan, and the United States indicated that the release from the Chernobyl reactor accident in the Soviet Union on April 26, 1986 contained a wide spectrum of fission up to heights of 7 km or more within a few days after the initial explosion. This high-altitude presence of radioactivity would in part be attributable to atmospheric dynamics factors other than the thermal energy released in the initial explosion. Indications were that two types of releases had taken place -- an initial powerful explosion followed by days of a less energetic reactor fire. The Atmospheric Release Advisory Capability (ARAC) at the Lawrence Livermore National Laboratory (LLNL) utilized three-dimensional atmospheric dispersion models to determine the characteristics of the source term (release) and the evolution of the spatial distributions of the airborne radioactivity as it was transported over Europe and subsequently over the northern hemisphere. This paper describes the ARAC involvement and the results of the hemispheric model calculations which graphically depict the extensive dispersal of radioactivity. 1 fig.

  4. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    NASA Astrophysics Data System (ADS)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  5. Simulation of impact of the Generic Accident-Resistant Packaging (GAP)

    SciTech Connect

    Slavin, A.M.

    1994-10-01

    Finite element simulations modelling impact of the Generic Accident-Resistant Packaging (GAP) have been performed. The GAP is a nuclear weapon shipping container that will be used by accident response groups from both the United States and the United Kingdom. The package is a thin-walled steel structure filled with rigid polyurethane foam and weighs approximately 5100 lbs when loaded. The simulations examined 250 ft/s impacts onto a rigid target at several orientations. The development of the finite element model included studies of modelling assumptions and material parameters. Upon completion of the simulation series, three full-scale impact tests were performed. A comparison of the simulation results to the test data is given. Differences between the results and data are examined, and possible explanations for the differences are discussed.

  6. Modular high-temperature gas-cooled reactor core heatup accident simulations

    SciTech Connect

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.

  7. Radiative transfer during the reflooding step of a LOCA

    NASA Astrophysics Data System (ADS)

    Gérardin, J.; Seiler, N.; Ruyer, P.; Boulet, P.

    2013-10-01

    Within the evaluation of the heat transfer downstream a quench front during the reflood phase of a Loss of Coolant Accident (LOCA) in a nuclear power plant, a numerical study has been conducted on radiative transfer through a vapor-droplet medium. The non-grey behavior of the medium is obvious since it can be optically thin or thick depending on the wavelength. A six wide bands model has been tested, providing a satisfactory accuracy for the description of the radiative properties. Once the radiative properties of the medium computed, they have been introduced in a model solving the radiative heat transfer based on the Improved Differential Approximation. The fluxes and the flux divergence have been computed on a geometry characteristic of the reactor core showing that radiative transfer plays a relevant role, quite as important as convective heat transfer.

  8. 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

  9. Fuel behavior during a LOCA: LOFT experiments

    SciTech Connect

    Russell, M.L.

    1980-11-01

    The LOFT experiments have provided the following fuel behavior information which appears to be valuable for improving the safety of PWR operation and resolving PWR licensing issues: (1) A generic unassisted core cooling event occurs during large-break LOCAs that dominates the cooling of the core before ECC reflood commences and potentially eliminates the possibility of flow channel blockage from prepressurized fuel rod swelling. (2) The large-break LOCA decompression forces do not disturb the normal control rod gravity drop and may not structually damage the fuel assemblies. (3) Large-break LOCA core cooling may also be enhanced by spacer grid and core counter flow delay of liquid escape from the core boundaries and liquid fallback from the upper plenum into the core region. (4) Lower fuel rod prepressurization may be possible in PWR fuel rods which would reduce flow channel blockage complications during LOCA's. (5) Uniform fuel rod cladding temperature indications during the large break LOCA's do not confirm expectations for the fuel rod cladding temperature variations that would inhibit development of flow channel blockages by ballooning of prepressurized fuel rods.

  10. Estimating the subjective risks of driving simulator accidents.

    PubMed

    Dixit, Vinayak; Harrison, Glenn W; Rutström, E Elisabet

    2014-01-01

    We examine the subjective risks of driving behavior using a controlled virtual reality experiment. Use of a driving simulator allows us to observe choices over risky alternatives that are presented to the individual in a naturalistic manner, with many of the cues one would find in the field. However, the use of a simulator allows us the type of controls one expects from a laboratory environment. The subject was tasked with making a left-hand turn into incoming traffic, and the experimenter controlled the headways of oncoming traffic. Subjects were rewarded for making a successful turn, and lost income if they crashed. The experimental design provided opportunities for subjects to develop subjective beliefs about when it would be safe to turn, and it also elicited their attitudes towards risk. A simple structural model explains behavior, and showed evidence of heterogeneity in both the subjective beliefs that subjects formed and their risk attitudes. We find that subjective beliefs change with experience in the task and the driver's skill. A significant difference was observed in the perceived probability to successfully turn among the inexperienced drivers who did and did not crash even though there was no significant difference in drivers' risk attitudes among the two groups. We use experimental economics to design controlled, incentive compatible tasks that provide an opportunity to evaluate the impact on driver safety of subject's subjective beliefs about when it would be safe to turn as well as their attitudes towards risk. This method could be used to help insurance companies determine risk premia associated with risk attitudes or beliefs of crashing, to better incentivize safe driving.

  11. Spatializing Sexuality in Jaime Hernandez's "Locas"

    ERIC Educational Resources Information Center

    Jones, Jessica E.

    2009-01-01

    Focusing on Jaime Hernandez's "Locas: The Maggie and Hopey Stories," part of the "Love and Rockets" comic series, I argue that the graphic landscape of this understudied comic offers an illustration of the theories of space in relation to race, gender, and sexuality that have been critical to understandings of Chicana sexuality. Set in a barrio…

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

    SciTech Connect

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

    2014-10-01

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

  13. Simulation Modeling Requirements for Loss-of-Control Accident Prevention of Turboprop Transport Aircraft

    NASA Technical Reports Server (NTRS)

    Crider, Dennis; Foster, John V.

    2012-01-01

    In-flight loss of control remains the leading contributor to aviation accident fatalities, with stall upsets being the leading causal factor. The February 12, 2009. Colgan Air, Inc., Continental Express flight 3407 accident outside Buffalo, New York, brought this issue to the forefront of public consciousness and resulted in recommendations from the National Transportation Safety Board to conduct training that incorporates stalls that are fully developed and develop simulator standards to support such training. In 2010, Congress responded to this accident with Public Law 11-216 (Section 208), which mandates full stall training for Part 121 flight operations. Efforts are currently in progress to develop recommendations on implementation of stall training for airline pilots. The International Committee on Aviation Training in Extended Envelopes (ICATEE) is currently defining simulator fidelity standards that will be necessary for effective stall training. These recommendations will apply to all civil transport aircraft including straight-wing turboprop aircraft. Government-funded research over the previous decade provides a strong foundation for stall/post-stall simulation for swept-wing, conventional tail jets to respond to this mandate, but turboprops present additional and unique modeling challenges. First among these challenges is the effect of power, which can provide enhanced flow attachment behind the propellers. Furthermore, turboprops tend to operate for longer periods in an environment more susceptible to ice. As a result, there have been a significant number of turboprop accidents as a result of the early (lower angle of attack) stalls in icing. The vulnerability of turboprop configurations to icing has led to studies on ice accumulation and the resulting effects on flight behavior. Piloted simulations of these effects have highlighted the important training needs for recognition and mitigation of icing effects, including the reduction of stall margins

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

    SciTech Connect

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

    1994-06-01

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

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

    SciTech Connect

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

    2012-10-01

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

  16. 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.

  17. Electrochemical behavior of simulated debris from a severe accident using a molten salt system

    SciTech Connect

    Takahashi, Yuya; Nakamura, Hitoshi; Yamada, Akira; Mizuguchi, Koji; Fujita, Reiko

    2013-07-01

    In a severe nuclear accident, the fuel in the reactor may melt, forming debris, which contains a UO{sub 2}-ZrO{sub 2} stable oxide mixture and parts of the reactor, such as Zircaloy and iron components. Proper handling of the debris is a critically important issue. The debris does not have the same composition as spent fuel, and so it is impossible to apply conventional reprocessing technology directly. In this study, we successfully separated Zr and Fe from simulated debris using NaCl-KCl molten salt electrolysis, and we selectively recovered the Zr and Fe. The simulated debris was made from Zr, Fe, and CeO{sub 2}. The CeO{sub 2} was used for simulating stable UO{sub 2}-ZrO{sub 2}. With this approach, it should be possible to reduce the volume of the debris by recovering metals, which can then be treated as low level radioactive wastes.

  18. Reconstruction of the 1994 Pittsburgh Airplane Accident Using a Computer Simulation

    NASA Technical Reports Server (NTRS)

    Parks, Edwin K.; Bach, Ralph E., Jr.; Shin, Jae Ho

    1998-01-01

    On September 8, 1994, a Boeing 737-300 passenger airplane was on a downwind approach to the Pittsburgh International Airport at an altitude of 5000 feet above ground level (6000 feet MSL). While in a shallow left turn onto a downwind approach heading, the airplane crossed into the vortex trail of a Boeing 727 flying in the same approach pattern about 4 miles ahead. The B-737 airplane rolled and turned sharply to the left, exited the vortex wake and plunged into the ground. Weather was not a factor in the accident. The airplane was equipped with a 11+ channel digital Flight Data Recorder (FDR) and a multiple channel Cockpit Voice Recorder (CVR). Both recorders were recovered from the crash site and provided excellent data for the development of an accident scenario. Radar tracking of the two airplanes as well as the indicated air speed (IAS) perturbations clearly visible on the B-737 FDR recordings indicate that the upset was apparently initiated by the airplane's crossing into the wake of the B-727 flying ahead in the same traffic pattern. A 6 degree-of-freedom simulation program for the B-737 airplane using MATLAB and SIMULINK was constructed. The simulation was initialized at the stabilized flight conditions of the airplane about 13 seconds prior to its entry into the vortex trail of the B-727 airplane. By assuming a certain combination of control inputs, it was possible to produce a simulated motion that closely matched that recorded on the FDR.

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

    SciTech Connect

    Alammar, M.A.

    1996-12-31

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

  20. Validation of the CALSPAN gross-motion-simulation code with actually occurring injury patterns in aircraft accidents.

    PubMed

    Ballo, J M; Dunne, M J; McMeekin, R R

    1978-01-01

    Digital simulation of aircraft-accident kinematics has heretofore been used almost exclusively as a design tool to explore structural load limits, precalculate decelerative forces at various cabin stations, and describe the effect of protective devices in the crash environment. In an effort to determine the value of digital computer simulation of fatal aircraft accidents, a fatality involving an ejection-system failure (out-of-envelope ejection) was modeled, and the injuries actually incurred were compared to those predicted; good agreement was found. The simulation of fatal aircraft accidents is advantageous because of a well-defined endpoint (death), lack of therapeutic intervention, and a static anatomic situation that can be minutely investigated. Such simulation techniques are a useful tool in the study of experimental trauma. PMID:623585

  1. Experimental investigation on the causes for pellet fragmentation under LOCA conditions

    NASA Astrophysics Data System (ADS)

    Bianco, A.; Vitanza, C.; Seidl, M.; Wensauer, A.; Faber, W.; Macián-Juan, R.

    2015-10-01

    This paper addresses a separate effect experiment performed with irradiated fuel to study fuel fragmentation and fission gas release during a loss of coolant accident (LOCA). The paper presents a qualitative and quantitative investigation of the effects of the removal of the geometrical constraint provided by the cladding and the removal of the constraint given by the rod internal pressure in determining the extent of fuel fragmentation and fission gas release during a LOCA for fuel segments with a burnup of approximately 52 MWd/kgU. A review of previous LOCA tests was the starting point for the identification of these constraints and for the selection of the fuel rod burnup, the experiment's procedure and the boundary conditions. An out-of-pile test was considered representative for the scope, and the experiment was performed at the Halden Reactor Project hot cell in Kjeller (Norway) with heat provided by an electric oven. Three fuel rod segments were studied: 1) a fuel segment that experienced only ballooning without burst, 2) a fuel segment that experienced ballooning and burst and 3) a fuel segment that experienced neither ballooning nor burst. The neutron radiography and fuel fragment sifting showed that both cladding constraint and internal pressure play a role in the formation of fuel cracks and fragmentation, and the study of the fission gas release during the transient showed that removing the cladding constraint or the internal pressure increased the amount of fission gas release.

  2. MADYMO simulation of children in cycle accidents: a novel approach in risk assessment.

    PubMed

    McNally, D S; Rosenberg, N M

    2013-10-01

    Head injuries are a significant cause of death and injury to child cyclists both on and off the road. Current evaluations of the effectiveness of cycle helmets rely on simplified mechanical testing or the analysis of aggregated accident statistics. This paper presents a direct evaluation of helmet efficacy by using computational modelling to simulate a range of realistic accident scenarios, including loss of control, collision with static objects and vehicle impact. A 6-year-old cyclist was modelled (as a Hybrid III 6-year-old dummy), in addition to a typical children's bicycle and a vehicle using the MADYMO dynamics software package. Simulations were performed using ranges of cyclist position, cycle speed and vehicle speed with and without a helmet that meets current standards. Wearing a cycle helmet was found to reduce the probability of head injuries, reducing the average probability of fatality over the scenarios studied from 40% to 0.3%. Similarly, helmet wearing reduced the probability of neck injuries (average probability of fatality reduced from 11% to 1%). There was no evidence that helmet wearing increased the severity of brain or neck injuries caused by rotational accelerations; in fact these were slightly reduced. Similarly, there was no evidence that increased cycling speed, such as might result from helmet related risk compensation, increased the probability of head injury. PMID:23954681

  3. Overview of Fuel Rod Simulator Usage at ORNL

    NASA Astrophysics Data System (ADS)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

  4. Overview of Fuel Rod Simulator Usage at ORNL

    SciTech Connect

    Ott, Larry J.; McCulloch, Reg

    2004-02-04

    During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

  5. An interactive framework for developing simulation models of hospital accident and emergency services.

    PubMed

    Codrington-Virtue, Anthony; Whittlestone, Paul; Kelly, John; Chaussalet, Thierry

    2005-01-01

    Discrete-event simulation can be a valuable tool in modelling health care systems. This paper describes an interactive framework to model and simulate a hospital accident and emergency department. An interactive spreadsheet (Excel) facilitated the user-friendly input of data such as patient pathways, arrival times, service times and resources into the discrete event simulation package (SIMUL8). The framework was enhanced further by configuring SIMUL8 to visually show patient flow and activity on a schematic plan of an A&E. The patient flow and activity information included patient icons flowing along A&E corridors and pathways, processes undertaken in A&E work areas and queue activity. One major benefit of visually showing patient flow and activity was that modellers and decision makers could visually gain a dynamic insight into the performance of the overall system and visually see changes over the model run cycle. Another key benefit of the interactive framework was the ability to quickly and easily change model parameters to trial, test and compare different scenarios.

  6. 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)

  7. Study of injuries combining computer simulation in motorcycle-car collision accidents.

    PubMed

    Guo, Lei; Jin, Xian-Long; Zhang, Xiao-Yun; Shen, Jie; Chen, Yi-Jiu; Chen, Jian-Guo

    2008-05-20

    This paper presents the approach of computer simulation to clarify the questions faced by forensic experts about what causes the various injuries characteristic of two motorcycle victims, including the motorcycle driver and the back seat occupant on the motorcycle, and how to exactly confirm which one of them is the motorcycle driver. Two typical motorcycle-car accident cases were reconstructed to analyze the movement and the load of both the motorcycle driver and the back seat occupant in the collision course. In case one, the back seat occupant suffered fatal head injuries when he fell on the ground after being thrown higher than the motorcycle driver over the top of the car. In case two, the compressive force loaded by the right tibia of the back seat occupant was larger and more durative compared with the motorcycle driver; the back seat occupant suffered a bursting fracture injury of his right tibia. These results might be useful for forensic experts in dealing with similar motorcycle-car collision accidents in the future.

  8. 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.

  9. A fast long-range transport model for operational use in episode simulation. Application to the Chernobyl accident

    NASA Astrophysics Data System (ADS)

    Bonelli, P.; Calori, G.; Finzi, G.

    A simple Lagrangian puff trajectory model and its software implementation, STRALE, are described. Standard meteorological data are used as input for the simulation of the three-dimensional atmospheric transport and dispersion of a pollutant released by a point source. The schemes adopted to describe the vertical diffusion and the interaction with the mixing layer are discussed on the basis of the comparison between simulated and measured 137Cs activities for the Chernobyl nuclear accident.

  10. 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)

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

    SciTech Connect

    Ball, S.J. )

    1991-10-01

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

  12. Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

    SciTech Connect

    Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

    2016-01-01

    Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the

  13. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in

  14. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

  15. Cellular automata model simulating traffic car accidents in the on-ramp system

    NASA Astrophysics Data System (ADS)

    Echab, H.; Lakouari, N.; Ez-Zahraouy, H.; Benyoussef, A.

    2015-01-01

    In this paper, using Nagel-Schreckenberg model we study the on-ramp system under the expanded open boundary condition. The phase diagram of the two-lane on-ramp system is computed. It is found that the expanded left boundary insertion strategy enhances the flow in the on-ramp lane. Furthermore, we have studied the probability of the occurrence of car accidents. We distinguish two types of car accidents: the accident at the on-ramp site (Prc) and the rear-end accident in the main road (Pac). It is shown that car accidents at the on-ramp site are more likely to occur when traffic is free on road A. However, the rear-end accidents begin to occur above a critical injecting rate αc1. The influence of the on-ramp length (LB) and position (xC0) on the car accidents probabilities is studied. We found that large LB or xC0 causes an important decrease of the probability Prc. However, only large xC0 provokes an increase of the probability Pac. The effect of the stochastic randomization is also computed.

  16. Long-term simulations of the 137Cs dispersion from the Fukushima accident in the world ocean.

    PubMed

    Nakano, Masanao; Povinec, Pavel P

    2012-09-01

    The LAMER calculation code was used for simulation of the distribution of (137)Cs released after the Fukushima accident into the Pacific and Indian Oceans. The maximum (137)Cs concentration predicted for surface waters of the open NW Pacific Ocean in 2012 (21 Bq/m(3) at 38 °N, 164 °E) will be comparable to that observed during the early 1960s after atmospheric nuclear weapons tests. The (137)Cs in surface waters of the Pacific Ocean will reach the US coast 4-5 y after the accident, however, the levels will be low (<3 Bq/m(3)). All the North Pacific Ocean will be labeled with Fukushima (137)Cs 10 y after the accident with concentrations below 1 Bq/m(3). Thirty years after the accident the (137)Cs levels in the Pacific and Indian Oceans will be below 0.1 Bq/m(3), i.e. undetectable on the present global fallout background. The effective dose commitment with ingestion of marine biota found in 2012 in the open NW Pacific Ocean was estimated to be 1.7 μSv/y, mostly delivered by (134,137)Cs. The estimated dose is by about a factor of 500 lower than the present dose limit for the public. PMID:22244453

  17. Long-term simulations of the 137Cs dispersion from the Fukushima accident in the world ocean.

    PubMed

    Nakano, Masanao; Povinec, Pavel P

    2012-09-01

    The LAMER calculation code was used for simulation of the distribution of (137)Cs released after the Fukushima accident into the Pacific and Indian Oceans. The maximum (137)Cs concentration predicted for surface waters of the open NW Pacific Ocean in 2012 (21 Bq/m(3) at 38 °N, 164 °E) will be comparable to that observed during the early 1960s after atmospheric nuclear weapons tests. The (137)Cs in surface waters of the Pacific Ocean will reach the US coast 4-5 y after the accident, however, the levels will be low (<3 Bq/m(3)). All the North Pacific Ocean will be labeled with Fukushima (137)Cs 10 y after the accident with concentrations below 1 Bq/m(3). Thirty years after the accident the (137)Cs levels in the Pacific and Indian Oceans will be below 0.1 Bq/m(3), i.e. undetectable on the present global fallout background. The effective dose commitment with ingestion of marine biota found in 2012 in the open NW Pacific Ocean was estimated to be 1.7 μSv/y, mostly delivered by (134,137)Cs. The estimated dose is by about a factor of 500 lower than the present dose limit for the public.

  18. 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.

  19. Modeling of Zr alloy burst cladding internal oxidation and secondary hydriding under LOCA conditions

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Shestak, V. E.

    2015-06-01

    The recently developed mechanistic model for Zr alloy cladding hydriding has been implemented in the single-rod SVECHA/QUENCH (S/Q) code. The mass transfer in a fuel rod after ballooning and burst opening have been modeled in the modified code that allowed calculating hydrogen and oxygen pickup by the cladding inner-metal surface. The code predicts with a good accuracy the typical distributions of oxygen and hydrogen in the Zr alloy cladding that were observed in the JAERI (Japan Atomic Energy Research Institute) and ANL (Argonne National Laboratory) single-rod tests and KIT (Karlsruhe Institute of Technology) bundle tests under postulated loss-of-coolant accident (LOCA) conditions.

  20. Simulation study of car accidents at the intersection of two roads in the mixed traffic flow

    NASA Astrophysics Data System (ADS)

    Marzoug, R.; Ez-Zahraouy, H.; Benyoussef, A.

    2015-05-01

    Using cellular automata (CA) Nagel-Schreckenberg (NaSch) model, we numerically study the probability Pac of the occurrence of car accidents at nonsignalized intersection when drivers do not respect the priority rules. We also investigated the impact of mixture lengths and velocities of vehicles on this probability. It is found that in the first case, where vehicles distinguished only by their lengths, the car accidents start to occur above a critical density ρc. Furthermore, the increase of the fraction of long vehicles (FL) delays the occurrence of car accidents (increasing ρc) and increases the risk of collisions when ρ > ρc. In other side, the mixture of maximum velocities (with same length for all vehicles) leads to the appearance of accidents at the intersection even in the free flow regime. Moreover, the increase of the fraction of fast vehicles (Ff) reduces the accident probability (Pac). The influence of roads length is also studied. We found that the decrease of the roads length enhance the risk of collision.

  1. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    SciTech Connect

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.

  2. Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

    SciTech Connect

    Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.; Phillips, Jesse; Kalinich, Donald A.; Osborn, Douglas M.; Peko, Damian

    2013-11-01

    Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

  3. Simulation study of traffic car accidents in single-lane highway

    NASA Astrophysics Data System (ADS)

    Bentaleb, Khalid; Lakouari, Noureddine; Marzoug, Rachid; Ez-Zahraouy, Hamid; Benyoussef, Abdelilah

    2014-11-01

    In this paper we numerically study the probability Pac of the occurrence of car accidents in the extended Nagel-Schreckenberg (NS) model in the case of mixture of fast (Vmax1=5) and slow vehicles (Vmax2=1) by taking also to the risky overtaking of fast vehicles. In comparison with previous existing models, we find that accidents can occur in the free traffic phase and/or congested one depending on the overtaking rate of fast vehicles. The effect of evacuation of damaged vehicles from the road with probabilities Pevf and Pevs of fast and slow vehicles respectively on the traffic flow behavior is also computed.

  4. Code System for Supercritical Water Cooled Reactor LOCA Analysis.

    1999-10-13

    Version 00 The new SCRELA code was developed to analyze the LOCA of the supercritical water cooled reactor. Since the currently available LWR codes for LOCA analysis could not analyze the significant differences in reactor characteristics between the supercritical-water cooled reactor and the current LWR, the first objective of this code development was to analyze the uniqueness of this reactor. The behavior of the supercritical water in the blowdown phase and the reflood phase ismore » modeled.« less

  5. Accident simulation and consequence analysis in support of MHTGR safety evaluations

    SciTech Connect

    Ball, S.J.; Wichner, R.P.; Smith, O.L.; Conklin, J.C. ); Barthold, W.P. )

    1991-01-01

    This paper summarizes research performed at Oak Ridge National Laboratory (ORNL) to assist the Nuclear Regulatory Commission (NRC) in preliminary determinations of licensability of the US Department of Energy (DOE) reference design of a standard modular high-temperature gas-cooled reactor (MHTGR). The work described includes independent analyses of core heatup and steam ingress accidents, and the reviews and analyses of fuel performance and fission product transport technology.

  6. TRAC-PF1/MOD1 US/Japanese PWR conservative LOCA prediction

    SciTech Connect

    Gruen, G E; Fisher, J E

    1987-11-01

    This report documents the results of a 200%, double-ended, cold-leg-break, loss-of-coolant-accident (LOCA) calculation using the TRAC-PF1/MOD1 computer code. The reactor system represented a typical United States/Japanese pressurized water reactor with a 15 x 15 fuel bundle arrangement 12-ft long, four loops, and cold-leg Emergency Core Cooling (ECC) Systems. Conservation boundary and initial conditions were used. Reactor power was 102% of the 3250 MWt rated power, decay heat was set to 120% of American Nuclear Society Standard 5.1, highest core lifetime values for power peaking and fuel stored energy were used, and the LOCA occurred simultaneously with a loss of offsite power. Best estimate assumptions were used for the break flow model, fuel rod heat transfer and metal-water reaction correlations, and steady-state fuel temperature profiles. A flow blockage model, having the capability to account for the effects of cladding ballooning or rupturing, was not used. Except for these best estimate assumptions, the boundary and initial conditions were consistent with those used in licensing calculations. Maximum fuel rod temperatures were 1380 K (2020/sup 0/F) and 1040 K (1410/sup 0/F) on the hottest evaluation model rod and hottest best estimate rod, respectively. The high reported values or fuel cladding temperature were a direct consequence of the conservative boundary and initial conditions used for the calculation, primarily the 2% overpower condition, the core decay heat assumption, and the degraded ECCS. The calculation demonstrated successful core reflooding before 1478 K (2200/sup 0/F) cladding temperature was exceeded on any fuel rod. 7 refs., 47 figs., 5 tabs.

  7. Best-Estimate Analysis PWR LOCA.

    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

  8. Stochastic Plume Simulations for the Fukushima Accident and the Deep Water Horizon Oil Spill

    NASA Astrophysics Data System (ADS)

    Coelho, E.; Peggion, G.; Rowley, C.; Hogan, P.

    2012-04-01

    The Fukushima Dai-ichi power plant suffered damage leading to radioactive contamination of coastal waters. Major issues in characterizing the extent of the affected waters were a poor knowledge of the radiation released to the coastal waters and the rather complex coastal dynamics of the region, not deterministically captured by the available prediction systems. Equivalently, during the Gulf of Mexico Deep Water Horizon oil platform accident in April 2010, significant amounts of oil and gas were released from the ocean floor. For this case, issues in mapping and predicting the extent of the affected waters in real-time were a poor knowledge of the actual amounts of oil reaching the surface and the fact that coastal dynamics over the region were not deterministically captured by the available prediction systems. To assess the ocean regions and times that were most likely affected by these accidents while capturing the above sources of uncertainty, ensembles of the Navy Coastal Ocean Model (NCOM) were configured over the two regions (NE Japan and Northern Gulf of Mexico). For the Fukushima case tracers were released on each ensemble member; their locations at each instant provided reference positions of water volumes where the signature of water released from the plant could be found. For the Deep Water Horizon oil spill case each ensemble member was coupled with a diffusion-advection solution to estimate possible scenarios of oil concentrations using perturbed estimates of the released amounts as the source terms at the surface. Stochastic plumes were then defined using a Risk Assessment Code (RAC) analysis that associates a number from 1 to 5 to each grid point, determined by the likelihood of having tracer particle within short ranges (for the Fukushima case), hence defining the high risk areas and those recommended for monitoring. For the Oil Spill case the RAC codes were determined by the likelihood of reaching oil concentrations as defined in the Bonn Agreement

  9. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    SciTech Connect

    Waeckel, N.; Cauvin, R.; Lebuffe, C.

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  10. Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs. Revision 7/95

    SciTech Connect

    Nourbakhsh, H.P.; Cheng, Z.

    1995-07-01

    This paper presents the results of a study of mixing phenomena related to boron dilution during small break loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler-condenser mode. A problem may occur when subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core are examined in this paper. Bounding calculations for boron concentration of coolant entering the core during a small break LOCA in a typical Westinghouse-designed four-loop plant are also presented.

  11. Simplified data assimilation for simulating wet deposition distribution of radioactive materials in FDNPP accident

    NASA Astrophysics Data System (ADS)

    Saya, A.; Yoshikane, T.; Chang, E. C.; Yoshimura, K.

    2015-12-01

    Due to the massive earthquakes and tsunami on March 11th 2011 in Eastern Japan, Fukushima Daiichi Nuclear Power Plant (FDNPP) was severely damaged. Radioactive materials were released and spread out by atmospheric advection-diffusion. Especially on March 21 - 23th when precipitation was observed, "hotspot" where the high concentration was detected locally. This area was formed in the metropolitan area in Kanto region. Thus, pollution at water treatment plants because of the deposition became a concern. Therefore, the reliable information of the hotspot is expected. Currently, atmospheric transport simulations by numerical models are developed for reproduction of the distribution. However, there are some uncertainties in the simulations. In the case of hotspot, accuracy of simulated precipitation have to be well considered because the hotspot seemed to be formed by wet deposition. We modified the stable isotope mode of Regional Spectral Model (IsoRSM) to enable to simulate the transport of the radioactive tracers, namely 131I and 137Cs, by including the dry and wet deposition processes. As the simplified data assimilation, simulated precipitation was replaced with Radar-AMeDAS precipitation data (RAP). RAP was assimilated in the post-process, after running simulations, to redistribute wet deposition of 137Cs. The ratio of 137Cs deposited from the cumulative vertical column with precipitation in the domain was not changed, however its pattern was redistributed corresponding with RAP and simulated concentration. As a result, the redistributed wet deposition was within factor 10 to 2 compared with the fallout data in Kanto region, and further data assimilation would be contributed. In addition, we found that due to the arrival time of the plume in the morning on 21st and the border time of daily observation data of fallout, validation result might be worse even though hourly distributions are well simulated.

  12. Molecular Dynamic Simulation of Sodium in 7-Pin LMFBR Bundle Under Hypothetical Accident Conditions

    SciTech Connect

    Bottoni, Maurizio; Bottoni, Claudio; Scanu, John

    2006-07-01

    In the frame of safety analysis of liquid metal fast breeder reactors (LMFBRs) under hypothetical Unprotected Loss of Flow (ULOF) conditions two-phase flow of sodium is simulated in a 7-pin bundle, with hexagonal lattice. Molecular dynamics, with the application of the Direct Simulation Monte Carlo (DSMC) method, and a macroscopic model describing rewetting sequences due to the flow of a sodium liquid film along the pin surfaces, are applied to simulate the coolant in the bundle. The pin surfaces and the inner surface of the hexagonal canning are treated in the Monte Carlo simulation as diffusively reflecting surfaces. Collisions of sodium molecules are computed with the 'hard-sphere' model. With respect to previous work the following improvements of the computational code were made: i) The full bundle is simulated, thus allowing for asymmetries, like a skewed power distribution, to be accounted for; ii) A pin model calculates detailed temperature distributions in the pins, so that temperature boundary conditions are computed and not imposed; iii) Post processing visualisation of computed results was developed. An out of pile sodium boiling experiment run at the Nuclear Research Center of Karlsruhe, Germany, is simulated and conclusions are drawn about the applicability of the methodology in computer codes dedicated to breeder reactors safety analysis. (authors)

  13. 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)

  14. For Whom the Bell Tolls: Imagining Accidents and the Development of Crisis Simulation in Organizations

    ERIC Educational Resources Information Center

    Smith, Denis

    2004-01-01

    This article explores how organizations can prepare for crisis events by training crisis management teams (CMTs) using real-time, simulated crises. The article focuses on the impact of such training on the performance of CMTs and the manner in which such training can improve the capability of the organization to deal with adverse events. The…

  15. Simulation of ¹³⁷Cs transport and deposition after the Chernobyl Nuclear Power Plant accident and radiological doses over the Anatolian peninsula.

    PubMed

    Simsek, V; Pozzoli, L; Unal, A; Kindap, T; Karaca, M

    2014-11-15

    The Chernobyl Nuclear Power Plant (CNPP) accident occurred on April 26 of 1986, it is still an episode of interest, due to the large amount of radionuclides dispersed in the atmosphere. Caesium-137 ((137)Cs) is one of the main radionuclides emitted during the Chernobyl accident, with a half-life of 30years, which can be accumulated in humans and animals, and for this reason the impacts on population are still monitored today. One of the main parameters in order to estimate the exposure of population to (137)Cs is the concentration in the air, during the days after the accident, and the deposition at surface. The transport and deposition of (137)Cs over Europe occurred after the CNPP accident has been simulated using the WRF-HYSPLIT modeling system. Four different vertical and temporal emission rate profiles have been simulated, as well as two different dry deposition velocities. The model simulations could reproduce fairly well the observations of (137)Cs concentrations and deposition, which were used to generate the 'Atlas of Caesium deposition on Europe after the Chernobyl accident' and published in 1998. An additional focus was given on (137)Cs deposition and air concentrations over Turkey, which was one of the main affected countries, but not included in the results of the Atlas. We estimated a total deposition of 2-3.5 PBq over Turkey, with 2 main regions affected, East Turkey and Central Black Sea coast until Central Anatolia, with values between 10 kBq m(-2) and 100 kBq m(-2). Mean radiological effective doses from simulated air concentrations and deposition has been estimated for Turkey reaching 0.15 mSv/year in the North Eastern part of Turkey, even if the contribution from ingestion of contaminated food and water is not considered, the estimated levels are largely below the 1 mSv limit indicated by the International Commission on Radiological Protection.

  16. 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.

  17. No longer waiting for an accident to happen: Simulation in emergency medicine.

    PubMed

    Ellison, Stefanie; Sullivan, Christine; McCullough, Robert

    2013-01-01

    The practice of emergency medicine (EM) requires proficient and expert skills in multiple high risk procedures. The emergency physician in-training needs a safe and realistic environment in which to practice and perfect the skills necessary to care for patients ranging from the critically ill to the patient with difficult intravenous access. Undergraduate medical, education overall has a need for training that enables students to develop the knowledge, skills and attitudes to practice in a variety of specialties. This article provides an overview of simulation in a three-year emergency medicine residency at Truman Medical Center, in a required final year clerkship for all medical students at the University of Missouri-Kansas City, and discusses national trends for the use of simulation in emergency medicine. PMID:23724485

  18. Effect of spray parameter on containment depressurization during LOCA in KAPP 3 and 4, 700 MWE IPHWR

    SciTech Connect

    Sharma, S. K.; Bhartia, D. K.; Mohan, N.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    KAPP 3 and 4 is an Indian Pressurized Heavy Water Reactor (IPHWR) of 700 MWe capacities. It is a pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 392 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. Containment of Indian PHWR is an ultimate barrier, which is designed to envelope whole reactor systems, to prevent the spread of active air-borne fission products in accident condition. Containment Spray System has been provided for energy as well as activity removal from the Containment system. This paper discusses about the studies done to assess the effect of spray parameters such as spray flow rate, droplets diameter and height of fall on containment peak pressure and temperature, long term containment depressurization and energy removal from the containment during Loss of Coolant Accident (LOCA). The spray flow rate and droplets diameter play an important role in removing residual energy from containment atmosphere, which influences depressurization of containment. It is obvious that faster depressurization of containment during postulated LOCA helps in limiting radiological consequences. From radiological considerations, droplets diameter is required to be kept to the lowest practically possible value and flow rate of spray should be high. Spray water droplets fall height governs the exposure time of droplets, which is the direct indication of energy removal rate. However, it is observed from the sensitivity studies that for a height of spray droplet fall more than 16.5 m, for the range of spray water flow rate and droplets sizes considered in the analyses, there is no significant change in heat removal. (authors)

  19. Core cooling under accident conditions at the high-flux beam reactor

    SciTech Connect

    Tichler, P.; Cheng, L. ); Fauske, H. )

    1991-01-01

    The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.

  20. 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.

  1. 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.

  2. 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.

  3. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  4. DSMC simulations in support of the Columbia Shuttle Orbiter accident investigation.

    SciTech Connect

    Gallis, Michail A.; Boyles, Katie A.; LeBeau, Gerald J.

    2004-06-01

    Three-dimensional Direct Simulation Monte Carlo simulations of Columbia Shuttle Orbiter flight STS-107 are presented. The aim of this work is to determine the aerodynamic and heating behavior of the Orbiter during aerobraking maneuvers and to provide piecewise integration of key scenario events to assess the plausibility of the candidate failure scenarios. The flight of the Orbiter is examined at two altitudes: 350-kft and 300-kft. The flowfield around the Orbiter and the heat transfer to it are calculated for the undamaged configuration. The flow inside the wing for an assumed damage to the leading edge in the form of a 10- inch hole is studied. The tragic loss of the Space Shuttle Columbia and her seven-member crew was followed by an investigation that lasted almost 7 months covering numerous failure scenarios. Due to the lack of physical data about flight STS-107 (especially in the high altitude part of it), numerical simulations were employed to help with the interpretation of the forensic evidence and the evaluation of the plausibility of the candidate scenarios. The conclusion of the investigation was that the physical cause of the loss of Columbia and its crew was a breach in the Thermal Protection System. To protect the aluminum structure of the Orbiter during re-entry, the Orbiter is covered with various materials collectively referred to as the Thermal Protection System. The three major components of the system are various types of heat-resistant tiles, blankets, and the Reinforced Carbon-Carbon (RCC) panels. The RCC panels are layers of graphite molded to the desired shape at very high temperatures. RCC is used for the Orbiter nose cap, chin panel, forward external tank attachment point, and wing leading edge panels and T-seals. RCC is a material capable of withstanding temperatures up to 2,000 K. Each wing leading edge consists of 22 RCC panels numbered from 1 to 22 moving outward on each wing. Because the shape of the wing changes from inboard to

  5. One-year, regional-scale simulation of 137Cs radioactivity in the ocean following the Fukushima Dai-ichi Nuclear Power Plant accident

    NASA Astrophysics Data System (ADS)

    Tsumune, D.; Tsubono, T.; Aoyama, M.; Uematsu, M.; Misumi, K.; Maeda, Y.; Yoshida, Y.; Hayami, H.

    2013-08-01

    A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant following the Great East Japan Earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways: direct release from the accident site and atmospheric deposition. A 1 yr, regional-scale simulation of 137Cs activity in the ocean offshore of Fukushima was carried out, the sources of radioactivity being direct release, atmospheric deposition, and the inflow of 137Cs deposited into the ocean by atmospheric deposition outside the domain of the model. Direct releases of 137Cs were estimated for 1 yr after the accident by comparing simulated results and measured activities adjacent to the accident site. The contributions of each source were estimated by analysis of 131I/137Cs and 134Cs/137Cs activity ratios and comparisons between simulated results and measured activities of 137Cs. The estimated total amounts of directly released 131I, 137Cs, and 137Cs were 11.1 ± 2.2 PBq, 3.5 ± 0.7 PBq, and 3.6 ± 0.7 PBq, respectively. Simulated 137Cs activities attributable to direct release were in good agreement with measured 137Cs activities not only adjacent to the accident site, but also in a wide area in the model domain, therefore this implies that the estimated direct release rate was reasonable. Employment of improved nudging data by JCOPE2 improved both the offshore transport result and the reproducibility of 137Cs activities 30 km offshore. On the other hand, simulated 137Cs activities attributable to atmospheric deposition were low compared to measured activities. The rate of atmospheric deposition into the ocean was underestimated because of a lack of measurements of deposition into the ocean when atmospheric deposition rates were being estimated. Simulated 137Cs activities attributable to the inflow of 137Cs deposited into the ocean outside the domain of the model were in good agreement with measured activities in the open ocean within the model

  6. Development of the aerosol generation system for simulating the dry deposition behavior of radioaerosol emitted by the accident of FDNPP

    NASA Astrophysics Data System (ADS)

    Zhang, Z.

    2015-12-01

    A large amount of radioactivity was discharged by the accident of FDNPP. The long half-life radionuclide, 137Cs was transported through the atmosphere mainly as the aerosol form and deposited to the forests in Fukushima prefecture. After the dry deposition of the 137Cs, the foliar uptake process would occur. To evaluate environmental transfer of radionuclides, the dry deposition and following foliar uptake is very important. There are some pioneering studies for radionuclide foliar uptake with attaching the solution containing stable target element on the leaf, however, cesium oxide aerosols were used for these deposition study [1]. In the FDNPP case, 137Cs was transported in sulfate aerosol form [2], so the oxide aerosol behaviors could not represent the actual deposition behavior in this accident. For evaluation of whole behavior of 137Cs in vegetation system, fundamental data for deposition and uptake process of sulfate aerosol was desired. In this study, we developed aerosol generation system for simulating the dry deposition and the foliar uptake behaviors of aerosol in the different chemical constitutions. In this system, the method of aerosol generation based on the spray drying. Solution contained 137Cs was send to a nozzle by a syringe pump and spraying with a high speed air flow. The sprayed mist was generated in a chamber in the relatively high temperature. The solution in the mist was dried quickly, and micro size solid aerosols consisting 137Cs were generated. The aerosols were suctioned by an ejector and transported inside a tube by the dry air flow, then were directly blown onto the leaves. The experimental condition, such as the size of chamber, chamber temperature, solution flow rate, air flow rate and so on, were optimized. In the deposition experiment, the aerosols on leaves were observed by a SEM/EDX system and the deposition amount was evaluated by measuring the stable Cs remaining on leaf. In the presentation, we will discuss the detail

  7. Large-eddy simulation of turbulent winds during the Fukushima Daiichi Nuclear Power Plant accident by coupling with a meso-scale meteorological simulation model

    NASA Astrophysics Data System (ADS)

    Nakayama, H.; Takemi, T.; Nagai, H.

    2015-06-01

    A significant amount of radioactive material was accidentally discharged into the atmosphere from the Fukushima Dai-ichi Nuclear Power Plant from 12 March 2011, which produced high contaminated areas over a wide region in Japan. In conducting regional-scale atmospheric dispersion simulations, the computer-based nuclear emergency response system WSPEEDI-II developed by Japan Atomic Energy Agency was used. Because this system is driven by a meso-scale meteorological (MM) model, it is difficult to reproduce small-scale wind fluctuations due to the effects of local terrain variability and buildings within a nuclear facility that are not explicitly represented in MM models. In this study, we propose a computational approach to couple an LES-based CFD model with a MM model for detailed simulations of turbulent winds with buoyancy effects under real meteorological conditions using turbulent inflow technique. Compared to the simple measurement data, especially, the 10 min averaged wind directions of the LES differ by more than 30 degrees during some period of time. However, distribution patterns of wind speeds, directions, and potential temperature are similar to the MM data. This implies that our coupling technique has potential performance to provide detailed data on contaminated area in the nuclear accidents.

  8. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    SciTech Connect

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio.

  9. Cladding burst behavior of Fe-based alloys under LOCA

    DOE PAGES

    Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; Massey, Caleb P.

    2015-12-17

    Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

  10. Cladding burst behavior of Fe-based alloys under LOCA

    SciTech Connect

    Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; Massey, Caleb P.

    2015-12-17

    Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.

  11. 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.

  12. Early results from an experimental program to determine the behavior of containment piping penetration bellows subjected to severe accident conditions

    SciTech Connect

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Containment piping penetration bellows are an integral part of the pressure boundary in steel containments in the United States (US). Their purpose is to minimize loading on the containment shell caused by differential movement between the piping and the containment. This differential movement is typically caused by thermal gradients generated during startup and shutdown of the reactor, but can be caused by earthquake, a loss-of-coolant accident (LOCA), or ``severe`` accidents. In the event of a severe accident, the bellows would be subjected to pressure, temperature, and deflection well beyond the design basis. Most bellows are installed such that they would be subjected to elevated internal pressure, elevated temperature, axial compression, and lateral deflection during a severe accident. A few bellows would be subjected to external pressure and axial elongation, as well as elevated temperature and lateral deflection. The purpose of this experimental program is to examine the potential for leakage of containment bellows during a severe accident. The test series subjects bellows to various levels and combinations of internal pressure, elevated temperature, axial compression or elongation, and lateral deformation. The experiments are being conducted in two parts. For Part 1, all bellows specimens are tested in ``like-new`` condition, without regard for the possible degrading effect of corrosion that has been observed in some containment piping bellows in the US Part I testing, which included 13 bellows tests, has been completed. The second part of the experimental program, in which bellows are subjected to simulated corrosive environments prior to testing, has just just begun. The Part I experiments have shown that bellows in ``like-new`` condition can withstand elevated temperatures and pressures along with large deformations before leaking. In most cases, the like-new bellows were fully compressed without developing any leakage.

  13. TRACG prediction of gravity-driven cooling system response in the SBWR/GIST facility LOCA tests

    SciTech Connect

    Alamgir, M.; Andersen, J.G.M.; Yang, A.I.; Shiralkar, B.S. )

    1990-01-01

    General Electric (BE) Nuclear Energy has initiated work on technology programs in support of the advanced light water reactor (ALWR) plants under contract to the U.S. Department of Energy (DOE). Work has been performed under the advanced boiling water reactor (ABWT) design verification program and the simplified boiling water reactor (SBWR) program. The objective of the SBWR program is to develop the key features of a simplified reactor design. The gravity-driven cooling system (GDCS) is an important feature of the SBWR design. The main objectives of the GDCS test program at GE were to demonstrate the technical feasibility of the GDCS concept by performing a section-scaled integrated systems test of the SBWR design and to provide a data base to qualify the TRACG computer code for use in SBWR accident analysis. This paper describes the qualification of TRACG for GDCS applications. The calculational capability and analytical models of TRACG are tested by performing assessment analysis for five loss-of-coolant-accident (LOCA) tests in the GDCS Integrated Systems Test (GIST) facility. The results of the qualification comparisons are presented and TRACG application ranges are discussed.

  14. 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.

  15. 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.

  16. First Responders and Criticality Accidents

    SciTech Connect

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

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

  17. Radiotherapy Accidents

    NASA Astrophysics Data System (ADS)

    Mckenzie, Alan

    A major benefit of a Quality Assurance system in a radiotherapy centre is that it reduces the likelihood of an accident. For over 20 years I have been the interface in the UK between the Institute of Physics and Engineering in Medicine and the media — newspapers, radio and TV — and so I have learned about radiotherapy accidents from personal experience. In some cases, these accidents did not become public and so the hospital cannot be identified. Nevertheless, lessons are still being learned.

  18. Probabilistic based design rules for intersystem LOCAS in ABWR piping

    SciTech Connect

    Ware, A.G.; Wesley, D.A.

    1993-05-01

    A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures.

  19. Experimental study of head loss and filtration for LOCA debris

    SciTech Connect

    Rao, D.V.; Souto, F.J.

    1996-02-01

    A series of controlled experiments were conducted to obtain head loss and filtration characteristics of debris beds formed of NUKON{trademark} fibrous fragments, and obtain data to validate the semi-theoretical head loss model developed in NUREG/CR-6224. A thermally insulated closed-loop test set-up was used to conduct experiments using beds formed of fibers only and fibers intermixed with particulate debris. A total of three particulate mixes were used to simulate the particulate debris. The head loss data were obtained for theoretical fiber bed thicknesses of 0.125 inches to 4.0 inches; approach velocities of 0.15 to 1.5 ft/s; temperatures of 75 F and 125 F; and sludge-to-fiber nominal concentration ratios of 0 to 60. Concentration measurements obtained during the first flushing cycle were used to estimate the filtration efficiencies of the debris beds. For test conditions where the beds are fairly uniform, the head loss data were predictable within an acceptable accuracy range by the semi-theoretical model. The model was equally applicable for both pure fiber beds and the mixed beds. Typically the model over-predicted the head losses for very thin beds and for thin beds at high sludge-to-fiber mass ratios. This is attributable to the non-uniformity of such debris beds. In this range the correlation can be interpreted to provide upper bound estimates of head loss. This is pertinent for loss of coolant accidents in boiling water reactors.

  20. Progress in the studies of passive heat removal in the next European torus under accident conditions

    SciTech Connect

    Soria, A. ); Renda, V.; Papa, L. . Joint Research Centre); Fenoglio, F. )

    1989-12-01

    Within the framework of safety analysis for the next European torus, a decay heat hazards assessment is under way in Ispra. Undercooling accidents (loss-of-coolant and loss-of-flow accidents (LOCAs and LOFAs)) due to pump failure have been investigated assuming an automatic plasma shutdown in both cases. The passive heat removal mechanisms considered include radiation between components and residual cooling by the thermosyphon effect in the main cooling circuits. Conservative thermohydraulic calculations have been made to determine coolant velocity and temperature transients to avoid water boiling int he circuits. Results show that during a LOFA, water boiling can be avoided provided that the water inertia is large enough, and material melting temperatures are not reached during a LOCA.

  1. Behaviour of oceanic 137Cs following the Fukushima Daiichi Nuclear Power Plant accident for four years simulated numerically by a regional ocean model

    NASA Astrophysics Data System (ADS)

    Torn, M. S.; Koven, C. D.; Riley, W. J.; Zhu, B.; Hicks Pries, C.; Phillips, C. L.

    2014-12-01

    A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant (1F NPP) following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition.We reconstructed spatiotemporal variability of 137Cs activity in the regional ocean for four years by numerical model, such as a regional scale and the North Pacific scale oceanic dispersion models, an atmospheric transport model, a sediment transport model, a dynamic biological compartment model for marine biota and river runoff model. Direct release rate of 137Cs were estimated for four years after the accident by comparing simulated results and observed activities very close to the site. The estimated total amounts of directly release was 3.6±0.7 PBq. Directly release rate of 137Cs decreased exponentially with time by the end of December 2012 and then, was almost constant. Decrease rate were quite small after 2013. The daily release rate of 137Cs was estimated to be the order of magnitude of 1010 Bq/day by the end of March 2015. The activity of directly released 137Cs was detectable only in the coastal zone after December 2012. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable. There is no observed data of 137Cs activity in the ocean from 11 to 21 March 2011. Observed data of marine biota should reflect the history of 137Cs activity in this early period. We reconstructed the history of 137Cs activity in this early period by considering atmospheric deposition, river input, rain water runoff from the 1F NPP site. The comparisons between simulated 137Cs activity of marine biota by a dynamic biological compartment and observed data also suggest that simulated 137Cs activity attributable to atmospheric deposition was underestimated in this early period. The

  2. Behaviour of oceanic 137Cs following the Fukushima Daiichi Nuclear Power Plant accident for four years simulated numerically by a regional ocean model

    NASA Astrophysics Data System (ADS)

    Tsumune, D.; Tsubono, T.; Aoyama, M.; Misumi, K.; Tateda, Y.

    2015-12-01

    A series of accidents at the Fukushima Dai-ichi Nuclear Power Plant (1F NPP) following the earthquake and tsunami of 11 March 2011 resulted in the release of radioactive materials to the ocean by two major pathways, direct release from the accident site and atmospheric deposition.We reconstructed spatiotemporal variability of 137Cs activity in the regional ocean for four years by numerical model, such as a regional scale and the North Pacific scale oceanic dispersion models, an atmospheric transport model, a sediment transport model, a dynamic biological compartment model for marine biota and river runoff model. Direct release rate of 137Cs were estimated for four years after the accident by comparing simulated results and observed activities very close to the site. The estimated total amounts of directly release was 3.6±0.7 PBq. Directly release rate of 137Cs decreased exponentially with time by the end of December 2012 and then, was almost constant. Decrease rate were quite small after 2013. The daily release rate of 137Cs was estimated to be the order of magnitude of 1010 Bq/day by the end of March 2015. The activity of directly released 137Cs was detectable only in the coastal zone after December 2012. Simulated 137Cs activities attributable to direct release were in good agreement with observed activities, a result that implies the estimated direct release rate was reasonable. There is no observed data of 137Cs activity in the ocean from 11 to 21 March 2011. Observed data of marine biota should reflect the history of 137Cs activity in this early period. We reconstructed the history of 137Cs activity in this early period by considering atmospheric deposition, river input, rain water runoff from the 1F NPP site. The comparisons between simulated 137Cs activity of marine biota by a dynamic biological compartment and observed data also suggest that simulated 137Cs activity attributable to atmospheric deposition was underestimated in this early period. The

  3. 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.

  4. 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.

  5. 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.

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

    DOE PAGES

    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

  7. Simulation of reflooding on two parallel heated channel by TRACE

    NASA Astrophysics Data System (ADS)

    Zakir, Md. Ghulam

    2016-07-01

    In case of Loss-Of-Coolant accident (LOCA) in a Boiling Water Reactor (BWR), heat generated in the nuclear fuel is not adequately removed because of the decrease of the coolant mass flow rate in the reactor core. This fact leads to an increase of the fuel temperature that can cause damage to the core and leakage of the radioactive fission products. In order to reflood the core and to discontinue the increase of temperature, an Emergency Core Cooling System (ECCS) delivers water under this kind of conditions. This study is an investigation of how the power distribution between two channels can affect the process of reflooding when the emergency water is injected from the top of the channels. The peak cladding temperature (PCT) on LOCA transient for different axial level is determined as well. A thermal-hydraulic system code TRACE has been used. A TRACE model of the two heated channels has been developed, and three hypothetical cases with different power distributions have been studied. Later, a comparison between a simulated and experimental data has been shown as well.

  8. Simulations of the transport and deposition of 137Cs over Europe after the Chernobyl NPP accident: influence of varying emission-altitude and model horizontal and vertical resolution

    NASA Astrophysics Data System (ADS)

    Evangeliou, N.; Balkanski, Y.; Cozic, A.; Møller, A. P.

    2013-03-01

    The coupled model LMDzORINCA has been used to simulate the transport, wet and dry deposition of the radioactive tracer 137Cs after accidental releases. For that reason, two horizontal resolutions were deployed and used in the model, a regular grid of 2.5°×1.25°, and the same grid stretched over Europe to reach a resolution of 0.45°×0.51°. The vertical dimension is represented with two different resolutions, 19 and 39 levels, respectively, extending up to mesopause. Four different simulations are presented in this work; the first uses the regular grid over 19 vertical levels assuming that the emissions took place at the surface (RG19L(S)), the second also uses the regular grid over 19 vertical levels but realistic source injection heights (RG19L); in the third resolution the grid is regular and the vertical resolution 39 vertical levels (RG39L) and finally, it is extended to the stretched grid with 19 vertical levels (Z19L). The best choice for the model validation was the Chernobyl accident which occurred in Ukraine (ex-USSR) on 26 May 1986. This accident has been widely studied since 1986, and a large database has been created containing measurements of atmospheric activity concentration and total cumulative deposition for 137Cs from most of the European countries. According to the results, the performance of the model to predict the transport and deposition of the radioactive tracer was efficient and accurate presenting low biases in activity concentrations and deposition inventories, despite the large uncertainties on the intensity of the source released. However, the best agreement with observations was obtained using the highest horizontal resolution of the model (Z19L run). The model managed to predict the radioactive contamination in most of the European regions (similar to Atlas), and also the arrival times of the radioactive fallout. As regards to the vertical resolution, the largest biases were obtained for the 39 layers run due to the increase of

  9. Atmospheric dispersion and ground deposition induced by the Fukushima Nuclear Power Plant accident: A local-scale simulation and sensitivity study

    NASA Astrophysics Data System (ADS)

    Korsakissok, I.; Mathieu, A.; Didier, D.

    2013-05-01

    Following the Fukushima Daiichi Nuclear Power Plant (FNPP1) accident on March 2011, radioactive products were released in the atmosphere. Simulations at local scale (within 80 km of FNPP1) were carried out by the Institute of Radiation Protection and Nuclear Safety (IRSN) with the Gaussian Puff model pX, during the crisis and since then, to assess the radiological and environmental consequences. The evolution of atmospheric and ground activity simulated at local scale is presented with a “reference” simulation, whose performance is assessed through comparisons with environmental monitoring data (gamma dose rate and deposition). The results are within a factor of 2-5 of the observations for gamma dose rates (0.52 and 0.85 for FAC2 and FAC5), and 5-10 for deposition (0.31 for FAC2, 0.73 for FAC5 and 0.90 for FAC10). A sensitivity analysis is also made to highlight the most sensitive parameters. A source term comparison is made between IRSN's estimation, and those from Katata et al. (2012) and Stohl et al. (2011). Results are quite sensitive to the source term, but also to wind direction and dispersion parameters. Dry deposition budget is more sensitive than wet deposition. Gamma dose rates are more sensitive than deposition, in particular peak values.

  10. RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

    SciTech Connect

    Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S.

    1995-09-01

    Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

  11. 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.

  12. World Meteorological Organization's model simulations of the radionuclide dispersion and deposition from the Fukushima Daiichi nuclear power plant accident.

    PubMed

    Draxler, Roland; Arnold, Dèlia; Chino, Masamichi; Galmarini, Stefano; Hort, Matthew; Jones, Andrew; Leadbetter, Susan; Malo, Alain; Maurer, Christian; Rolph, Glenn; Saito, Kazuo; Servranckx, René; Shimbori, Toshiki; Solazzo, Efisio; Wotawa, Gerhard

    2015-01-01

    Five different atmospheric transport and dispersion model's (ATDM) deposition and air concentration results for atmospheric releases from the Fukushima Daiichi nuclear power plant accident were evaluated over Japan using regional (137)Cs deposition measurements and (137)Cs and (131)I air concentration time series at one location about 110 km from the plant. Some of the ATDMs used the same and others different meteorological data consistent with their normal operating practices. There were four global meteorological analyses data sets available and two regional high-resolution analyses. Not all of the ATDMs were able to use all of the meteorological data combinations. The ATDMs were configured identically as much as possible with respect to the release duration, release height, concentration grid size, and averaging time. However, each ATDM retained its unique treatment of the vertical velocity field and the wet and dry deposition, one of the largest uncertainties in these calculations. There were 18 ATDM-meteorology combinations available for evaluation. The deposition results showed that even when using the same meteorological analysis, each ATDM can produce quite different deposition patterns. The better calculations in terms of both deposition and air concentration were associated with the smoother ATDM deposition patterns. The best model with respect to the deposition was not always the best model with respect to air concentrations. The use of high-resolution mesoscale analyses improved ATDM performance; however, high-resolution precipitation analyses did not improve ATDM predictions. Although some ATDMs could be identified as better performers for either deposition or air concentration calculations, overall, the ensemble mean of a subset of better performing members provided more consistent results for both types of calculations. PMID:24182910

  13. World Meteorological Organization's model simulations of the radionuclide dispersion and deposition from the Fukushima Daiichi nuclear power plant accident.

    PubMed

    Draxler, Roland; Arnold, Dèlia; Chino, Masamichi; Galmarini, Stefano; Hort, Matthew; Jones, Andrew; Leadbetter, Susan; Malo, Alain; Maurer, Christian; Rolph, Glenn; Saito, Kazuo; Servranckx, René; Shimbori, Toshiki; Solazzo, Efisio; Wotawa, Gerhard

    2015-01-01

    Five different atmospheric transport and dispersion model's (ATDM) deposition and air concentration results for atmospheric releases from the Fukushima Daiichi nuclear power plant accident were evaluated over Japan using regional (137)Cs deposition measurements and (137)Cs and (131)I air concentration time series at one location about 110 km from the plant. Some of the ATDMs used the same and others different meteorological data consistent with their normal operating practices. There were four global meteorological analyses data sets available and two regional high-resolution analyses. Not all of the ATDMs were able to use all of the meteorological data combinations. The ATDMs were configured identically as much as possible with respect to the release duration, release height, concentration grid size, and averaging time. However, each ATDM retained its unique treatment of the vertical velocity field and the wet and dry deposition, one of the largest uncertainties in these calculations. There were 18 ATDM-meteorology combinations available for evaluation. The deposition results showed that even when using the same meteorological analysis, each ATDM can produce quite different deposition patterns. The better calculations in terms of both deposition and air concentration were associated with the smoother ATDM deposition patterns. The best model with respect to the deposition was not always the best model with respect to air concentrations. The use of high-resolution mesoscale analyses improved ATDM performance; however, high-resolution precipitation analyses did not improve ATDM predictions. Although some ATDMs could be identified as better performers for either deposition or air concentration calculations, overall, the ensemble mean of a subset of better performing members provided more consistent results for both types of calculations.

  14. Downscaling humidity with Localized Constructed Analogs (LOCA) over the conterminous United States

    NASA Astrophysics Data System (ADS)

    Pierce, D. W.; Cayan, D. R.

    2016-07-01

    Humidity is important to climate impacts in hydrology, agriculture, ecology, energy demand, and human health and comfort. Nonetheless humidity is not available in some widely-used archives of statistically downscaled climate projections for the western U.S. In this work the Localized Constructed Analogs (LOCA) statistical downscaling method is used to downscale specific humidity to a 1°/16° grid over the conterminous U.S. and the results compared to observations. LOCA reproduces observed monthly climatological values with a mean error of ~0.5 % and RMS error of ~2 %. Extreme (1-day in 1- and 20-years) maximum values (relevant to human health and energy demand) are within ~5 % of observed, while extreme minimum values (relevant to agriculture and wildfire) are within ~15 %. The asymmetry between extreme maximum and minimum errors is largely due to residual errors in the bias correction of extreme minimum values. The temporal standard deviations of downscaled daily specific humidity values have a mean error of ~1 % and RMS error of ~3 %. LOCA increases spatial coherence in the final downscaled field by ~13 %, but the downscaled coherence depends on the spatial coherence in the data being downscaled, which is not addressed by bias correction. Temporal correlations between daily, monthly, and annual time series of the original and downscaled data typically yield values >0.98. LOCA captures the observed correlations between temperature and specific humidity even when the two are downscaled independently.

  15. Status of (137)Cs contamination in marine biota along the Pacific coast of eastern Japan derived from a dynamic biological model two years simulation following the Fukushima accident.

    PubMed

    Tateda, Yutaka; Tsumune, Daisuke; Tsubono, Takaki; Misumi, Kazuhiro; Yamada, Masatoshi; Kanda, Jota; Ishimaru, Takashi

    2016-01-01

    Radiocesium ((134)Cs and (137)Cs) released into the Fukushima coastal environment was transferred to marine biota inhabiting the Pacific Ocean coastal waters of eastern Japan. Though the levels in most of the edible marine species decreased overtime, radiocesium concentrations in some fishes were still remained higher than the Japanese regulatory limit for seafood products. In this study, a dynamic food chain transfer model was applied to reconstruct (137)Cs levels in olive flounder by adopting the radiocesium concentrations in small demersal fish which constitute an important fraction of the diet of the olive flounder particularly inhabiting area near Fukushima. In addition, (137)Cs levels in slime flounder were also simulated using reported radiocesium concentrations in some prey organisms. The simulated results from Onahama on the southern border of the Fukushima coastline, and at Choshi the southernmost point where the contaminated water mass was transported by the Oyashio current, were assessed in order to identify what can be explained from present information, and what remains to be clarified three years after the Fukushima Dai-ichi nuclear power plant (1FNPP) accident. As a result, the observed (137)Cs concentrations in planktivorous fish and their predator fish could be explained by the theoretically-derived simulated levels. On the other hand, the slow (137)Cs depuration in slime flounder can be attributed to uptake from unknown sources for which the uptake fluxes were of a similar magnitude as the excretion fluxes. Since the reported (137)Cs concentrations in benthic invertebrates off Onahama were higher than the simulated values, radiocesium transfer from these benthic detritivorous invertebrates to slime flounder via ingestion was suggested as a cause for the observed slow depuration of (137)Cs in demersal fish off southern Fukushima. Furthermore, the slower depuration in the demersal fish likely required an additional source of (137)Cs, i

  16. 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)

  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. 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

  19. Documentation for Three Wake Vortex Model Data Sets from Simulation of Flight 587 Wake Vortex Encounter Accident Case

    NASA Technical Reports Server (NTRS)

    Switzer, George F.

    2008-01-01

    This document contains a general description for data sets of a wake vortex system in a turbulent environment. The turbulence and thermal stratification of the environment are representative of the conditions on November 12, 2001 near John F. Kennedy International Airport. The simulation assumes no ambient winds. The full three dimensional simulation of the wake vortex system from a Boeing 747 predicts vortex circulation levels at 80% of their initial value at the time of the proposed vortex encounter. The linked vortex oval orientation showed no twisting, and the oval elevations at the widest point were about 20 meters higher than where the vortex pair joined. Fred Proctor of NASA?s Langley Research Center presented the results from this work at the NTSB public hearing that started 29 October 2002. This document contains a description of each data set including: variables, coordinate system, data format, and sample plots. Also included are instructions on how to read the data.

  20. Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

    SciTech Connect

    Ratnayake, Ruwan K.; Ergun, S.; Hochreiter, L.E.; Baratta, A.J.

    2002-07-01

    In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and 5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena. (authors)

  1. 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

  2. Large break LOCA calculations for the AP600 design

    SciTech Connect

    Fisher, J.E.

    1992-08-01

    This paper presents the application of RELAP5 to the calculation of a Large Break (200% doubled-ended rupture) Loss-of-Colant-Accident (LBLOCA) at the reactor vessel inlet for the proposed Westinghouse AP600 design. A parametric calculation was also performed to determine effects of loss of a complete Emergency Core Cooling system (ECCS) train. These calculations were performed over the core blowdown, refill, and reflood phases of the LBLOCA and did not address long term cooling. RELAP5 was shown to be adequate for system response calculation over the period of interest. The passive safety systems were predicted to effectively mitigate the consequences of LBLOCAs; the calculations showed less severe thermal responses than for a current generation Pressurized Water Reactor (PWR) plant. The two primary differences between the AP600 design and a current generation plant that affect LBLOCA response are the lower core thermal power, which results in lower temperatures during the blowdown phase, and the long duration accumulator injection, which provides ample core inventory makeup for final quenching.

  3. 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.

  4. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 1: STEALTH 1D single-phase fluid studies. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Mortensen, G.A.; Caraher, D.L.

    1980-04-01

    This report, which is the first in a series of reports for RP-1065, describes the first step in the stepwise approach for developing the methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The first step in the methodology considers enhancements and special modifications to the 1D STEALTH computer code in order that acoustic phenomena in piping and vessel networks may be simulated. The resulting code is termed 1D STEALTH-HYDRO. The 1D STEALTH-HYDRO enhancements consist of three control volume models to simulate area changes, orifices, and tees in piping networks. The theory of the control volume models is described.

  5. 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.

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

    SciTech Connect

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

    1994-10-01

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

  7. LOCA hydroloads calculations with multidimensional nonlinear fluid/structure interaction. Volume 3. Fluid/structure interaction studies using 3-D STEALTH/WHAMSE. Final report. [PWR

    SciTech Connect

    Santee, G.E. Jr.; Chang, F.H.; Mortensen, G.A.; Brockett, G.F.; Gross, M.B.; Belytschko, T.B.

    1982-11-01

    This report, the third in a series of reports for RP-1065, describes the final step in the stepwise approach for developing the three-dimensional, nonlinear, fluid-structure interaction methodology to assess the hydroloads on a large PWR during the subcooled portions of a hypothetical LOCA. The final step in the methodology implements enhancements and special modifications to the STEALTH 3D computer program and the WHAMSE 3D computer program. After describing the enhancements, the individual and the coupled computer programs are assessed by comparing calculational results with either analytical solutions or with experimental data. The coupled 3D STEALTH/WHAMSE computer program is then applied to the simulation of HDR Test V31.1 to further assess the program and to investigate the role that fluid-structure interaction plays in the hydrodynamic loading of reactor internals during subcooled blowdown.

  8. 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.

  9. The Accident Sequence Precursor program: Methods improvements and current results

    SciTech Connect

    Minarick, J.W.; Manning, F.M.; Harris, J.D.

    1987-01-01

    Changes in the US NRC Accident Sequence Precursor program methods since the initial program evaluations of 1969-81 operational events are described, along with insights from the review of 1984-85 events. For 1984-85, the number of significant precursors was consistent with the number observed in 1980-81, dominant sequences associated with significant events were reasonably consistent with PRA estimates for BWRs, but lacked the contribution due to small-break LOCAs previously observed and predicted in PWRs, and the frequency of initiating events and non-recoverable system failures exhibited some reduction compared to 1980-81. Operational events which provide information concerning additional PRA modeling needs are also described.

  10. Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

    NASA Astrophysics Data System (ADS)

    Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

    2014-06-01

    In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

  11. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    NASA Astrophysics Data System (ADS)

    Osadchaya, D. Yu.; Fuks, R. L.

    2014-04-01

    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  12. 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)

  13. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect

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

    2012-11-01

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

  14. 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.

  15. Comparison of TRAC and RELAP5 reactor system calculations for a DEGB LOCA in K-14.1

    SciTech Connect

    Griggs, D.P.; Liebmann, M.L.

    1990-09-01

    A comparison of TRAC and RELAP5 predictions of steady-state and DEGB LOCA results (FI phase) for K-14.1 has been made. Both codes had been previously benchmarked against 1985 L Reactor AC Flow data and were under configuration control. The purpose of the code-to-code comparison is to provide insight on the transient uncertainty in TRAC plenum and tank bottom plenum pressures. The comparisons focus on LOCA results between 0.5 and 2.0 s, which is the primary period of interest for Flow Instability (FI) limits.

  16. Comparison of TRAC and RELAP5 reactor system calculations for a DEGB LOCA in K-14. 1

    SciTech Connect

    Griggs, D.P. ); Liebmann, M.L. )

    1990-09-01

    A comparison of TRAC and RELAP5 predictions of steady-state and DEGB LOCA results (FI phase) for K-14.1 has been made. Both codes had been previously benchmarked against 1985 L Reactor AC Flow data and were under configuration control. The purpose of the code-to-code comparison is to provide insight on the transient uncertainty in TRAC plenum and tank bottom plenum pressures. The comparisons focus on LOCA results between 0.5 and 2.0 s, which is the primary period of interest for Flow Instability (FI) limits.

  17. Development and first application of a new tool for the simulation of the initiating phase of a severe accident on SFR

    NASA Astrophysics Data System (ADS)

    Guyot, M.; Gubernatis, P.; Suteau, C.

    2014-06-01

    In order to improve the safety level of Sodium Fast Reactors, low probability events such as Hypothetical Core Disruptive Accident (HCDA) are analyzed for their potential consequences. The initiating phase of such accidents is of particular interest both for the prevention and the mitigation of routes leading to a large core disruption and recriticalities. Up to now, analysis of the initiating phase of HCDA has been performed with the SAS4A code. The SAS4A accident calculations are based on a multiple-channel approach, which requires that subassemblies or groups of similar subassemblies be represented together as independent channels. The SAS4A severe accident calculation scheme resorts to a simplified treatment in which an average pin is used to represent a channel. A point kinetics model coupled with a feedback reactivity model is also used to provide an estimate of the reactor power level. Both to increase the accuracy and decrease the uncertainties in the prediction of reactor safety margins, a new computational tool is currently under development at CEA Cadarache. The main features of this tool are the ability to provide a detailed sub-channel meshing of the sub-assembly as well as three-dimensional kinetics during severe accident conditions. To fulfill these goals, the fluid-dynamics SIMMER-III code has been coupled to the SNATCH solver using a MPI environment. This coupling allows both to compute the multi-phase and multi-component flows encountered in severe accident conditions and to model the power shape variation during voiding and melting of the different reactor materials. This new calculation scheme relies on a SAS-like multiple-channel treatment, where channel-to-channel heat and momentum exchanges are neglected. In this paper, an overview of the SIMMER-III/SNATCH coupled tool capabilities is provided. A first application of this new tool is also performed and compared with a SAS4A reference calculation. The new SIMMER-III/SNATCH tool proved to be

  18. Continuation and bifurcation analysis of large-scale dynamical systems with LOCA.

    SciTech Connect

    Salinger, Andrew Gerhard; Phipps, Eric Todd; Pawlowski, Roger Patrick

    2010-06-01

    Dynamical systems theory provides a powerful framework for understanding the behavior of complex evolving systems. However applying these ideas to large-scale dynamical systems such as discretizations of multi-dimensional PDEs is challenging. Such systems can easily give rise to problems with billions of dynamical variables, requiring specialized numerical algorithms implemented on high performance computing architectures with thousands of processors. This talk will describe LOCA, the Library of Continuation Algorithms, a suite of scalable continuation and bifurcation tools optimized for these types of systems that is part of the Trilinos software collection. In particular, we will describe continuation and bifurcation analysis techniques designed for large-scale dynamical systems that are based on specialized parallel linear algebra methods for solving augmented linear systems. We will also discuss several other Trilinos tools providing nonlinear solvers (NOX), eigensolvers (Anasazi), iterative linear solvers (AztecOO and Belos), preconditioners (Ifpack, ML, Amesos) and parallel linear algebra data structures (Epetra and Tpetra) that LOCA can leverage for efficient and scalable analysis of large-scale dynamical systems.

  19. 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.

  20. Core cooling under accident conditions at the high flux beam reactor (HFBR)

    SciTech Connect

    Tichler, P.; Cheng, L. ); Fauske, H. )

    1991-01-01

    In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. In a LOCA scenario where even limited fuel damage occurs and natural circulation is established, fission product gases could be carried from the damaged fuel by steam into areas where operator access is required to maintain the core in a coolable configuration. This would force evacuation of the building and lead to extensive core damage. As a result the HFBR was shut down by the Department of Energy (DOE) and an extensive review of the HFBR was initiated. In an effort to address this issue BNL developed a model designed to predict the heat removal limit during flow reversal that was found to be in good agreement with the test results. Currently a thermal-hydraulic test program is being developed to provide a more realistic and defensible estimate of the flow reversal heat removal limit so that the reactor power level can be increased.

  1. A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

    SciTech Connect

    Brad J. Merrill; Richard L. Moore; J. Phillip Sharp

    2008-09-01

    A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium-steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.

  2. 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.

  3. Laser accidents: Being Prepared

    SciTech Connect

    Barat, K

    2003-01-24

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

  4. Simulations of the transport and deposition of 137Cs over Europe after the Chernobyl Nuclear Power Plant accident: influence of varying emission-altitude and model horizontal and vertical resolution

    NASA Astrophysics Data System (ADS)

    Evangeliou, N.; Balkanski, Y.; Cozic, A.; Møller, A. P.

    2013-07-01

    The coupled model LMDZORINCA has been used to simulate the transport, wet and dry deposition of the radioactive tracer 137Cs after accidental releases. For that reason, two horizontal resolutions were deployed and used in the model, a regular grid of 2.5° × 1.27°, and the same grid stretched over Europe to reach a resolution of 0.66° × 0.51°. The vertical dimension is represented with two different resolutions, 19 and 39 levels respectively, extending up to the mesopause. Four different simulations are presented in this work; the first uses the regular grid over 19 vertical levels assuming that the emissions took place at the surface (RG19L(S)), the second also uses the regular grid over 19 vertical levels but realistic source injection heights (RG19L); in the third resolution the grid is regular and the vertical resolution 39 levels (RG39L) and finally, it is extended to the stretched grid with 19 vertical levels (Z19L). The model is validated with the Chernobyl accident which occurred in Ukraine (ex-USSR) on 26 May 1986 using the emission inventory from Brandt et al. (2002). This accident has been widely studied since 1986, and a large database has been created containing measurements of atmospheric activity concentration and total cumulative deposition for 137Cs from most of the European countries. According to the results, the performance of the model to predict the transport and deposition of the radioactive tracer was efficient and accurate presenting low biases in activity concentrations and deposition inventories, despite the large uncertainties on the intensity of the source released. The best agreement with observations was obtained using the highest horizontal resolution of the model (Z19L run). The model managed to predict the radioactive contamination in most of the European regions (similar to De Cort et al., 1998), and also the arrival times of the radioactive fallout. As regards to the vertical resolution, the largest biases were obtained for

  5. A first look at LOCAs in the SBWR using RELAP5/MOD3

    SciTech Connect

    Ghan, L.S.; Shaw, R.A.; Kullberg, C.M.

    1992-01-01

    The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE's Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.

  6. A first look at LOCAs in the SBWR using RELAP5/MOD3

    SciTech Connect

    Ghan, L.S.; Shaw, R.A.; Kullberg, C.M.

    1992-12-31

    The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE`s Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.

  7. Civil aircraft accident investigation.

    PubMed

    Haines, Daniel

    2013-01-01

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

  8. Anatomy of an Accident.

    ERIC Educational Resources Information Center

    Mobley, Michael

    1984-01-01

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

  9. Farm accidents in children.

    PubMed Central

    Cameron, D.; Bishop, C.; Sibert, J. R.

    1992-01-01

    OBJECTIVE--To examine the problem of accidental injury to children on farms. DESIGN--Prospective county based study of children presenting to accident and emergency departments over 12 months with injuries sustained in a farm setting and nationwide review of fatal childhood farm accidents over the four years April 1986 to March 1990. SETTING--Accident and emergency departments in Aberystwyth, Carmarthen, Haverfordwest, and Llanelli and fatal accidents in England, Scotland, and Wales notified to the Health and Safety Executive register. SUBJECTS--Children aged under 16. MAIN OUTCOME MEASURE--Death or injury after farm related accidents. RESULTS--65 accidents were recorded, including 18 fractures. Nine accidents necessitated admission to hospital for a mean of two (range one to four) days. 13 incidents were related to tractors and other machinery; 24 were due to falls. None of these incidents were reported under the statutory notification scheme. 33 deaths were notified, eight related to tractors and allied machinery and 10 related to falling objects. CONCLUSIONS--Although safety is improving, the farm remains a dangerous environment for children. Enforcement of existing safety legislation with significant penalties and targeting of safety education will help reduce accident rates further. PMID:1638192

  10. 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.

  11. Persistence of airline accidents.

    PubMed

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

    2010-10-01

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

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

    SciTech Connect

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

    1988-05-01

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

  13. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    NASA Astrophysics Data System (ADS)

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2015-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of cumulative

  14. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    SciTech Connect

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C.

    2012-07-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  15. Accident resistant transport container

    DOEpatents

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

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

  16. Accident resistant transport container

    DOEpatents

    Andersen, John A.; Cole, James K.

    1980-01-01

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

  17. Safety Is No Accident.

    ERIC Educational Resources Information Center

    Christiansen, Monty L.

    1985-01-01

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

  18. Accident management information needs

    SciTech Connect

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

    1990-04-01

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

  19. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    NASA Astrophysics Data System (ADS)

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2014-06-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of

  20. A computer program for assessment of emergency operation procedures under non-loca transient conditions in BWRs

    SciTech Connect

    Ohga, Y.

    1983-06-01

    A program analyzing long-term transients after abnormal incidents, excluding loss-of-coolant accidents, has been developed to assess emergency operation procedures for cold shutdown of reactors. The main program features are: The thermal hydraulics in both the reactor pressure vessel (RPV) and the primary containment vessel (PCV) are treated. Analytical models of the cooling system are included for not only the emergency core cooling system but other cooling systems that are effective for RPV and PCV cooling. The on/off switching of cooling systems by plant interlocks, component failures, and operator actions is simulated. The applicability of this program has been evaluated by simulation of long-term thermal-hydraulic behavior of the boiling water reactor transients initiated by loss of feedwater. From the evaluation results, it has been confirmed that the main program models can assess emergency operation procedures.

  1. Image computing techniques to extrapolate data for dust tracking in case of an experimental accident simulation in a nuclear fusion plant.

    PubMed

    Camplani, M; Malizia, A; Gelfusa, M; Barbato, F; Antonelli, L; Poggi, L A; Ciparisse, J F; Salgado, L; Richetta, M; Gaudio, P

    2016-01-01

    In this paper, a preliminary shadowgraph-based analysis of dust particles re-suspension due to loss of vacuum accident (LOVA) in ITER-like nuclear fusion reactors has been presented. Dust particles are produced through different mechanisms in nuclear fusion devices, one of the main issues is that dust particles are capable of being re-suspended in case of events such as LOVA. Shadowgraph is based on an expanded collimated beam of light emitted by a laser or a lamp that emits light transversely compared to the flow field direction. In the STARDUST facility, the dust moves in the flow, and it causes variations of refractive index that can be detected by using a CCD camera. The STARDUST fast camera setup allows to detect and to track dust particles moving in the vessel and then to obtain information about the velocity field of dust mobilized. In particular, the acquired images are processed such that per each frame the moving dust particles are detected by applying a background subtraction technique based on the mixture of Gaussian algorithm. The obtained foreground masks are eventually filtered with morphological operations. Finally, a multi-object tracking algorithm is used to track the detected particles along the experiment. For each particle, a Kalman filter-based tracker is applied; the particles dynamic is described by taking into account position, velocity, and acceleration as state variable. The results demonstrate that it is possible to obtain dust particles' velocity field during LOVA by automatically processing the data obtained with the shadowgraph approach. PMID:26827318

  2. Image computing techniques to extrapolate data for dust tracking in case of an experimental accident simulation in a nuclear fusion plant

    NASA Astrophysics Data System (ADS)

    Camplani, M.; Malizia, A.; Gelfusa, M.; Barbato, F.; Antonelli, L.; Poggi, L. A.; Ciparisse, J. F.; Salgado, L.; Richetta, M.; Gaudio, P.

    2016-01-01

    In this paper, a preliminary shadowgraph-based analysis of dust particles re-suspension due to loss of vacuum accident (LOVA) in ITER-like nuclear fusion reactors has been presented. Dust particles are produced through different mechanisms in nuclear fusion devices, one of the main issues is that dust particles are capable of being re-suspended in case of events such as LOVA. Shadowgraph is based on an expanded collimated beam of light emitted by a laser or a lamp that emits light transversely compared to the flow field direction. In the STARDUST facility, the dust moves in the flow, and it causes variations of refractive index that can be detected by using a CCD camera. The STARDUST fast camera setup allows to detect and to track dust particles moving in the vessel and then to obtain information about the velocity field of dust mobilized. In particular, the acquired images are processed such that per each frame the moving dust particles are detected by applying a background subtraction technique based on the mixture of Gaussian algorithm. The obtained foreground masks are eventually filtered with morphological operations. Finally, a multi-object tracking algorithm is used to track the detected particles along the experiment. For each particle, a Kalman filter-based tracker is applied; the particles dynamic is described by taking into account position, velocity, and acceleration as state variable. The results demonstrate that it is possible to obtain dust particles' velocity field during LOVA by automatically processing the data obtained with the shadowgraph approach.

  3. Image computing techniques to extrapolate data for dust tracking in case of an experimental accident simulation in a nuclear fusion plant.

    PubMed

    Camplani, M; Malizia, A; Gelfusa, M; Barbato, F; Antonelli, L; Poggi, L A; Ciparisse, J F; Salgado, L; Richetta, M; Gaudio, P

    2016-01-01

    In this paper, a preliminary shadowgraph-based analysis of dust particles re-suspension due to loss of vacuum accident (LOVA) in ITER-like nuclear fusion reactors has been presented. Dust particles are produced through different mechanisms in nuclear fusion devices, one of the main issues is that dust particles are capable of being re-suspended in case of events such as LOVA. Shadowgraph is based on an expanded collimated beam of light emitted by a laser or a lamp that emits light transversely compared to the flow field direction. In the STARDUST facility, the dust moves in the flow, and it causes variations of refractive index that can be detected by using a CCD camera. The STARDUST fast camera setup allows to detect and to track dust particles moving in the vessel and then to obtain information about the velocity field of dust mobilized. In particular, the acquired images are processed such that per each frame the moving dust particles are detected by applying a background subtraction technique based on the mixture of Gaussian algorithm. The obtained foreground masks are eventually filtered with morphological operations. Finally, a multi-object tracking algorithm is used to track the detected particles along the experiment. For each particle, a Kalman filter-based tracker is applied; the particles dynamic is described by taking into account position, velocity, and acceleration as state variable. The results demonstrate that it is possible to obtain dust particles' velocity field during LOVA by automatically processing the data obtained with the shadowgraph approach.

  4. Injuries are not accidents

    PubMed Central

    Gutiérrez, María Isabel

    2014-01-01

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

  5. Accident prevention manual

    SciTech Connect

    Not Available

    1998-05-01

    Among the many common needs and goals are the safety and well-being of families, ourselves, fellow employees, and the continuing success of this organization. To these ends--minimizing human suffering and economic waste--the Bonneville Power Administration (BPA) Accident Prevention Program and this Accident Prevention Manual (APM) are dedicated. The BPA Accident Prevention Program is revised as necessary to ensure compliance with relevant Federal safety and health standards. The mandatory rules herein express minimum requirements for dealing with the principal hazards inherent in daily work activities. These and other written requirements, which neither can nor should provide complete coverage of all work situations, must be continually reinforced through the sound and mature safety judgments of all workers on each assigned task. In the event of conflicting judgments, the more conservative interpretation shall prevail pending review and resolution by management.

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

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

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

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

    PubMed

    Meineke, Viktor; Dörr, Harald

    2012-08-01

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

  8. 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.

  9. The generation of aerosols by accidents which may occur during plant-scale production of micro-organisms.

    PubMed Central

    Ashcroft, J.; Pomeroy, N. P.

    1983-01-01

    Experiments have been performed to simulate accidents which may occur during large-scale production of micro-organisms. Four types of accident, which were considered to be the most likely to result in the greatest hazard to health, were simulated using a bacterial model. The accidents were all concerned with faults occurring in the operation of the microbial fermenter. Gross contamination of surfaces occurred in all experiments, but only three types of accident produced a measurable aerosol. PMID:6350448

  10. Occupational accidents aboard merchant ships

    PubMed Central

    Hansen, H; Nielsen, D; Frydenberg, M

    2002-01-01

    Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were identified during a total of 31 140 years at sea. Among these, 209 accidents resulted in permanent disability of 5% or more, and 27 were fatal. The mean risk of having an occupational accident was 6.4/100 years at sea and the risk of an accident causing a permanent disability of 5% or more was 0.67/100 years aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded rate of accidents than Danish citizens. Age was a major risk factor for accidents causing permanent disability. Change of ship and the first period aboard a particular ship were identified as risk factors. Walking from one place to another aboard the ship caused serious accidents. The most serious accidents happened on deck. Conclusions: It was possible to clearly identify work situations and specific risk factors for accidents aboard merchant ships. Most accidents happened while performing daily routine duties. Preventive measures should focus on workplace instructions for all important functions aboard and also on the prevention of accidents caused by walking around aboard the ship. PMID:11850550

  11. Characterizing the Severe Turbulence Environments Associated with Commercial Aviation Accidents. Part 2; Hydrostatic Mesobeta Scale Numerical Simulations of Supergradient Wind Flow and Streamwise Ageostrophic Frontogenesis

    NASA Technical Reports Server (NTRS)

    Kaplan, Michael L.; Huffman, Allan W.; Lux, Kevin M.; Cetola, Jeffrey D.; Charney, Joseph J.; Riordan, Allen J.; Lin, Yuh-Lang; Waight, Kenneth T., III; Proctor, Fred (Technical Monitor)

    2003-01-01

    Simulation experiments reveal key processes that organize a hydrostatic environment conducive to severe turbulence. The paradigm requires juxtaposition of the entrance region of a curved jet stream, which is highly subgeostrophic, with the entrance region of a straight jet stream, which is highly supergeostrophic. The wind and mass fields become misphased as the entrance regions converge resulting in the significant spatial variation of inertial forcing, centripetal forcing, and along- and cross-stream pressure gradient forcing over a mesobeta scale region. This results in frontogenesis and the along-stream divergence of cyclonic and convergence of cyclonic ageostrophic vertical vorticity. The centripetally forced mesoscale front becomes the locus of large gradients of ageostrophic vertical vorticity along an overturning isentrope. This region becomes favorable for streamwise vorticity gradient formation enhancing the environment for organization of horizontal vortex tubes in the presence of buoyant forcing.

  12. 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

  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. Tractor accidents in Swedish traffic.

    PubMed

    Pinzke, Stefan; Nilsson, Kerstin; Lundqvist, Peter

    2012-01-01

    The objective of this study is to reach a better understanding of accidents on Swedish roads involving tractors and to suggest ways of preventing them. In an earlier study we analyzed police-reported fatal accidents and accidents that led to physical injuries from 1992 to 2005. During each year of this period, tractors were involved in 128 traffic accidents on average, an average of 7 people were killed, 44 sustained serious injuries, and 143 sustained slight injuries. The number of fatalities in these tractor accidents was about 1.3% of all deaths in traffic accidents in Sweden. Cars were most often involved in the tractor accidents (58%) and 15% were single vehicle accidents. The mean age of the tractor driver involved was 39.8 years and young drivers (15-24 years) were overrepresented (30%). We are now increasing the data collected with the years 2006-2010 in order to study the changes in the number of accidents. Special attention will be given to the younger drivers and to single vehicle accidents. Based on the results we aim to develop suggestions for reducing road accidents, e.g. including measures for making farm vehicles more visible and improvement of the training provided at driving schools. PMID:22317543

  15. Cooperation of mobile robots for accident scene inspection

    NASA Astrophysics Data System (ADS)

    Byrne, R. H.; Harrington, J.

    A telerobotic system demonstration was developed for the Department of Energy's Accident Response group to highlight the applications of telerobotic vehicles to accident site inspection. The proof-of-principle system employs two mobile robots, Dixie and RAYBOT, to inspect a simulated accident site. Both robots are controlled serially from a single driving station, allowing an operator to take advantage of having multiple robots at the scene. The telerobotic system is described and some of the advantages of having more than one robot present are discussed. Future plans for the system are also presented.

  16. 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.

  17. Post-accident inhalation exposure and experience with plutonium

    SciTech Connect

    Shinn, J

    1998-06-01

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

  18. [Skateboard and rollerskate accidents].

    PubMed

    Lohmann, M; Petersen, A O; Pedersen, O D

    1990-05-28

    The increasing popularity of skateboards and rollerskates has resulted in an increased number of contacts with the casualty department in Denmark after accidents. As part of the Danish share in the EHLASS project (European Home and Leisure Surveillance System), 120,000 consecutive contacts with the casualty departments were reviewed. Out of these 516 were due to accidents with skateboards and rollerskates (181/335). A total of 194 of these injuries (38%) were fractures and 80% of these were in the upper limbs. Twenty fractures required reposition under general anaesthesia and two required osteosynthesis. Nine patients were admitted for observation for concussion. One patient had sustained rupture of the spleen and splenectomy was necessary. A total of 44 patients were admitted. None of the 516 patients had employed protective equipment on the injured region. Considerable reduction in the number of injuries could probably be produced by employment of suitable protective equipment.

  19. [Drowning accidents in childhood].

    PubMed

    Krandick, G; Mantel, K

    1990-09-30

    This is a report on five boys aged between 1 and 5 years who, after prolonged submersion in cold water, were treated at our department. On being taken out of the water, all the patients were clinically dead. After 1- to 3-hour successful cardiopulmonary resuscitation, with a rectal temperature of about 27 degrees C, they were rewarmed at a rate of 1 degree/hour. Two patients died within a few hours after the accident. One patient survived with an apallic syndrome, 2 children survived with no sequelae. In the event of a water-related accident associated with hypothermia, we consider suitable resuscitation to have preference over rewarming measures. The most important treatment guidelines and prognostic factors are discussed.

  20. Farm accidents in children.

    PubMed

    Cogbill, T H; Busch, H M; Stiers, G R

    1985-10-01

    During a 6 1/2 year period, 105 children were admitted to the hospital as the result of trauma that occurred on farms. The mechanism of injury was animal related in 42 (40%), tractor or wagon accident in 28 (26%), farm machinery in 21 (20%), fall from farm building in six (6%), and miscellaneous in eight (8%). Injury Severity Score was calculated for each patient. An Injury Severity Score of greater than or equal to 25 was determined for 11 children (11%). Life-threatening injuries, therefore, are frequently the result of childhood activities that take place in agricultural environments. The most common injuries were orthopedic, neurologic, thoracoabdominal, and maxillofacial. There was one death in the series, and only one survivor sustained major long-term disability. Such injuries are managed with optimal outcome in a regional trauma center. Educational programs with an emphasis on prevention and safety measures may reduce the incidence of farm accidents. PMID:4047799

  1. Evaluation of prompt nucleation of bubbles in annular fuel elements during the initial depressurization transient of a DEGB LOCA

    SciTech Connect

    Smith, A.C.

    1997-06-01

    In the first moments following the pipe break, of a DEGB LOCA, the depressurization wave is postulated to propagate rapidly through the system, in the manner of an acoustic or water hammer wave. this is immediately followed by a (reflected) repressurization wave, as the flow of coolant through the break is established. The pressure history is then dictated by the flow from the break and the ability of the pressurizer, pumps and accumulators to supply coolant. The initial sudden drop in pressure may result in the system pressure falling below the saturation pressure of the coolant. This could, in turn, result in bubble formation. Such immediate vapor formation (prompt nucleation of bubbles), in the period before the repressurization wave restores the system pressure to a level above the saturation pressure might initiate flow instability. Such an interruption in flow would allow the fuel tube clad temperature to increase rapidly. Depending on the duration of the flow interruption, the reactor might not be able to survive the initial moments of DEGB LOCA. It has generally been that this phenomenon would not actually occur in an operating reactor. The purpose of this investigation is to evaluate the possibility of occurrence of bubble formation as a result of initial depressurization. 7 refs., 6 figs.

  2. Accident Flying Squad

    PubMed Central

    Snook, Roger

    1972-01-01

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

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

    PubMed

    Takada, Jun

    2012-03-01

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

  4. Aircraft accident survivors as witnesses.

    PubMed

    Dodge, R E

    1983-02-01

    This is a study of the reliability of aircrash survivors as witnesses. Some of their statements are compared to known facts at the time of the crash, including the time of the accident and the weather conditions. Other facts are compared between the survivors, such as the mood of the passengers immediately post-crash. The KLM-Pan Am accident in the Canary Islands is used as the study accident. A suggestion for future use of survivors' statements is tendered.

  5. Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

    SciTech Connect

    Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

    2002-07-01

    The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, {sup A}cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors{sup .} These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from

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

    PubMed Central

    Sehmer, J. M.

    1993-01-01

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

  7. Radiation accident grips Goiania

    SciTech Connect

    Roberts, L.

    1987-11-20

    On 13 September two young scavengers in Goiania, Brazil, removed a stainless steel cylinder from a cancer therapy machine in an abandoned clinic, touching off a radiation accident second only to Chernobyl in its severity. On 18 September they sold the cylinder, the size of a 1-gallon paint can, to a scrap dealer for $25. At the junk yard an employee dismantled the cylinder and pried open the platinum capsule inside to reveal a glowing blue salt-like substance - 1400 curies of cesium-137. Fascinated by the luminescent powder, several people took it home with them. Some children reportedly rubbed in on their bodies like carnival glitter - an eerie image of how wrong things can go when vigilance over radioactive materials lapses. In all, 244 people in Goiania, a city of 1 million in central Brazil, were contaminated. The eventual toll, in terms of cancer or genetic defects, cannot yet be estimated. Parts of the city are cordoned off as radiation teams continue washing down buildings and scooping up radioactive soil. The government is also grappling with the political fallout from the accident.

  8. German aircraft accident statistics, 1930

    NASA Technical Reports Server (NTRS)

    Weitzmann, Ludwig

    1932-01-01

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

  9. Transport aircraft accident dynamics

    NASA Technical Reports Server (NTRS)

    Cominsky, A.

    1982-01-01

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

  10. Risk Estimation Methodology for Launch Accidents.

    SciTech Connect

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

    2014-02-01

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

  11. 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.

  12. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User's guide

    SciTech Connect

    Rettig, W.H.; Wade, N.L. )

    1992-06-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  13. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis, Model description

    SciTech Connect

    Borkowski, J.A.; Wade, N.L.; Giles, M.M.; Rouhani, S.Z.; Shumway, R.W.; Singer, G.L.; Taylor, D.D.; Weaver, W.L. )

    1992-08-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODl version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MODI version produces results consistent with previous versions. Assessment calculations using the two TRAC-BF1 versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  14. Consequences of severe nuclear accidents in Europe

    NASA Astrophysics Data System (ADS)

    Seibert, Petra; Arnold, Delia; Mraz, Gabriele; Arnold, Nikolaus; Gufler, Klaus; Kromp-Kolb, Helga; Kromp, Wolfgang; Sutter, Philipp

    2013-04-01

    A first part of the presentation is devoted to the consequences of the severe accident in the 1986 Chernobyl NPP. It lead to a substantial radioactive contaminated of large parts of Europe and thus raised the awareness for off-site nuclear accident consequences. Spatial patterns of the (transient) contamination of the air and (persistent) contamination of the ground were studied by both measurements and model simulations. For a variety of reasons, ground contamination measurements have variability at a range of spatial scales. Results will be reviewed and discussed. Model simulations, including inverse modelling, have shown that the standard source term as defined in the ATMES study (1990) needs to be updated. Sensitive measurements of airborne activities still reveal the presence of low levels of airborne radiocaesium over the northern hemisphere which stems from resuspension. Over time scales of months and years, the distribution of radionuclides in the Earth system is constantly changing, for example relocated within plants, between plants and soil, in the soil, and into water bodies. Motivated by the permanent risk of transboundary impacts from potential major nuclear accidents, the multidisciplinary project flexRISK (see http://flexRISK.boku.ac.at) has been carried out from 2009 to 2012 in Austria to quantify such risks and hazards. An overview of methods and results of flexRISK is given as a second part of the presentation. For each of the 228 NPPs, severe accidents were identified together with relevant inventories, release fractions, and release frequencies. Then, Europe-wide dispersion and dose calculations were performed for 2788 cases, using the Lagrangian particle model FLEXPART. Maps of single-case results as well as various aggregated risk parameters were produced. It was found that substantial consequences (intervention measures) are possible for distances up to 500-1000 km, and occur more frequently for a distance range up to 100-300 km, which is in

  15. A review of criticality accidents

    SciTech Connect

    Stratton, W R; Smith, D R

    1989-03-01

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

  16. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    NASA Astrophysics Data System (ADS)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  17. Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

    SciTech Connect

    Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

    2004-10-15

    Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report.

  18. [Prevention of bicycle accidents].

    PubMed

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

    2015-04-01

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

  19. [Prevention of bicycle accidents].

    PubMed

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

    2015-04-01

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

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

    SciTech Connect

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

    2015-01-01

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

  1. [Orofacial injuries in skateboard accidents].

    PubMed

    Frohberg, U; Bonsmann, M

    1992-04-01

    In a clinical study, 25 accidents involving injuries by a fall with a skateboard were investigated and classified in respect of epidemiology, accident mechanism and injury patterns in the facial region. Accident victims are predominantly boys between 7 and 9 years of age. A multiple trauma involving the teeth and the dental system in general and the soft parts of the face is defined as a characteristic orofacial injury pattern in skateboard accidents. The high proportion of damage to the front teeth poses problems of functional and aesthetic rehabilitation necessitating long-term treatment courses in children and adolescents. Effective prevention of facial injuries may be possible by evolving better facial protection systems and by creating areas of playgrounds where skateboarders can practise safely.

  2. 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)

  3. Spine Immobilizer for Accident Victims

    NASA Technical Reports Server (NTRS)

    Vykukal, H. C.; Lampson, K.

    1983-01-01

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

  4. 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.

  5. 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

  6. [Accidents of fulguration].

    PubMed

    Virenque, C; Laguerre, J

    1976-01-01

    Fulguration, first electric accident in which the man was a victim, is to day better known. A clap of thunder is decomposed in two elements: lightning, and thunder. Lightning is caused by an electrical discharge, either within a cloud, or between two clouds, or, above all, between a cloud and the surface of the ground. Experimental equipments owned by the French Electricity Company and by the Atomic Energy Commission, have allowed to photograph lightnings and to measure certain physical characteristics (Intensity variable between 25 to 100 kA, voltage variable between 20 to 1 000 kV). The frequency of storms was learned: the isokeraunic level, in France, is about 20, meaning that thunder is heard twenty days during one year. Man may be stricken by thunder by direct hit, by sudden bursting, by earth current, or through various conductors. The electric charge which reached him may go to the earth directly by contact with the ground or may dissipate in the air through a bony promontory (elbow). The total number of victims, "wounded" or deceased, is not now known by statistics. Death comes by insulation breakdown of one of several anatomic cephalic formations: skull, meninx, brain. Many various lesions may happen in survivors: loss of consciousness, more or less long, sensorial or motion deficiencies. All these signs are momentary and generally reversible. Besides one may observe much more intense lesions on the skin: burns and, over all, characteristic aborescence (skin effect by high frequency current). The heart is protected, contrarily to what happens with industrial electrocution. The curative treatment is merely symptomatic : reanimation, surgery for burns or associated traumatic lesions. A prevention is researched to help the lonely man, in the country or in the mountains in the houses (lightning conductor, Faraday cage), in vehicles (aircraft, cars, ships). The mysterious and unforseeable character of lightning still stays, leaving a door opened for numerous

  7. 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

  8. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    SciTech Connect

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  9. 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.

  10. Post-quench ductility evaluation of Zircaloy-4 and select iron alloys under design basis and extended LOCA conditions

    NASA Astrophysics Data System (ADS)

    Yan, Y.; Keiser, J. R.; Terrani, K. A.; Bell, G. L.; Snead, L. L.

    2014-05-01

    Oxidation experiments were conducted at 1200 °C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post-quench ductility under postulated and extended LOCA conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 °C was determined to be k = 2.173 × 107 g2/cm4/s, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post-quenched samples was evaluated by ring compression tests at 135 °C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR ≈ 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to 4 h.

  11. TRACG Simulation of Drywell Gas Recirculation System in ESBWR

    SciTech Connect

    Cheung, Yee K.; Rao, Atambir S.

    2002-07-01

    This paper presents the results of a parametric study on the mitigating effects of the Drywell Gas Recirculation System (DGRS) in ESBWR during postulated LOCA and severe accidents. The post-accident containment pressure depends on the sum of the partial pressure from non-condensable gases and partial steam pressure inside the wet-well airspace. Results of parametric studies show that, with the activation of DGRS: (1) The containment pressure continues to reduce due to the redistribution of non-condensable gases from the wet-well back to the drywell; (2) The DGRS can be designed in a 'portable' fashion; (3) The Current ESBWR meets the design requirement with significant margin using only passive safety systems, and the margin increases considerably with the activation of DGRS. (authors)

  12. 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.

  13. Secondary hydriding during LOCA - Results from the QUENCH-L0 test

    NASA Astrophysics Data System (ADS)

    Grosse, Mirco; Stuckert, Juri; Steinbrück, Martin; Kaestner, Anders

    2012-01-01

    A shortened loss of coolant scenario was simulated experimentally at fuel rod bundle scale in the QUENCH-L0 test. All pressurized specimens burst at temperatures between 790 and 855 °C. The hydrogen distribution in the cladding of chosen fuel rod simulators was studied by neutron radiography and tomography. As the results of these investigations show, the inner oxidation level, depending on time between burst and cooling and the temperatures occurring during this time, are important parameters for the hydrogen uptake. No hydrogen uptake was measured for the specimen with less than 50 s between burst and quenching. For longer time intervals, hydrogen enriched bands were formed. These bended bands are oriented non-symmetrically to the tube axis. The width of the bands and the maximal hydrogen concentrations seem to increase with time between burst and quenching and with the temperatures during this period.

  14. Source term and radiological consequences of the Chernobyl accident

    SciTech Connect

    Mourad, R.; Snell, V.

    1987-01-01

    The objective of this work is to assess the source term and to evaluate the maximum hypothetical individual doses in European countries (including the Soviet Union) from the Chernobyl accident through the analyses of measurements of meteorological data, radiation fields, and airborne and deposited activity in these countries. Applying this information to deduce the source term involves a reversal of the techniques of nuclear accident analysis, which estimate the off-site consequences of postulated accidents. In this study the authors predict the quantities of radionuclides that, if released at Chernobyl and following the calculated trajectories, would explain and unify the observed radiation levels and radionuclide concentrations as measured by European countries and the Soviet Union. The simulation uses the PEAR microcomputer program following the methodology described in Canadian Standards Association standard N288.2. The study was performed before the Soviets published their estimate of the source term and the two results are compared.

  15. Identification of severe accident uncertainties

    SciTech Connect

    Rivard, J.B.; Behr, V.L.; Easterling, R.G.; Griesmeyer, J.M.; Haskin, F.E.; Hatch, S.W.; Kolaczkowski, A.M.; Lipinski, R.J.; Sherman, M.P.; Taig, A.R.

    1984-09-01

    Understanding of severe accidents in light-water reactors is currently beset with uncertainty. Because the uncertainties that are present limit the capability to analyze the progression and possible consequences of such accidents, they restrict the technical basis for regulatory actions by the US Nuclear Regulatory Commission (NRC). It is thus necessary to attempt to identify the sources and quantify the influence of these uncertainties. As a part of ongoing NRC severe-accident programs at Sandia National Laboratories, a working group was formed to pool relevant knowledge and experience in assessing the uncertainties attending present (1983) knowledge of severe accidents. This initial report of the Severe Accident Uncertainty Analysis (SAUNA) working group has as its main goal the identification of a consolidated list of uncertainties that affect in-plant processes and systems. Many uncertainties have been identified. A set of key uncertainties summarizes many of the identified uncertainties. Quantification of the influence of these uncertainties, a necessary second step, is not attempted in the present report, although attempts are made qualitatively to demonstrate the relevance of the identified uncertainties.

  16. The child accident repeater: a review.

    PubMed

    Jones, J G

    1980-04-01

    The child accident repeater is defined as one who has at least three accidents that come to medical attention within a year. The accident situation has features in common with those of the child who has a single accident through simple "bad luck", but other factors predispose him to repeated injury. In the child who has a susceptible personality, a tendency for accident repetition may be due to a breakdown in adjustment to a stressful environment. Prevention of repeat accidents should involve the usual measures considered appropriate for all children as well as an attempt to provide treatment of significant maladjustment and modification of a stressful environment.

  17. Investigation of the Challenger Accident

    NASA Technical Reports Server (NTRS)

    1986-01-01

    The work of the Presidential Commission on the Space Shuttle Challenger Accident (hereafter referred to as the Rogers Commission) and the work of the National Aeronautics and Space Administration in investigating the causes of the accident were reviewed. In addition to reviewing the five volumes of the Rogers Commission, the entire direct on-line Rogers Commission data base, which included full-text and document retrieval capability was also reviewed. The findings and recommendations contained also include materials submitted for the record, staff investigations, interviews, and trips.

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

    PubMed

    Schröder, S; Lier, H; Wiese, S

    2004-11-01

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

  19. Geochemical and Hydrologic Controls of Copper-Rich Surface Waters in the Yerba Loca-Mapocho System

    NASA Astrophysics Data System (ADS)

    Pasten, P.; Montecinos, M.; Coquery, M.; Pizarro, G. E.; Abarca, M. I.; Arce, G. J.

    2015-12-01

    Andean watersheds in Northern and Central Chile are naturally enriched with metals, many of them associated to sulfide mineralizations related to copper mining districts. The natural and anthropogenic influx of toxic metals into drinking water sources pose a sustainability challenge for cities that need to provide safe water with the smallest footprint. This work presents our study of the transformations of copper in the Yerba Loca-Mapocho system. Our sampling campaign started from the headwaters at La Paloma Glacier and continues to the inlet of the San Enrique drinking water treatment plant, a system feeding municipalities in the Eastern area of Santiago, Chile. Depending on the season, total copper concentrations go as high as 22 mg/L for the upper sections, which become diluted to <5 mg/L downstream. pH ranged from 3 to 5.6 while suspended solids ranged from <10 to 100 mg/L. We used Geochemist Workbench to assess copper speciation and to evaluate the thermodynamic controls for the formation and dissolution of solid phases. A sediment trap was used to concentrate suspended particulate matter, which was analyzed with ICP-MS, TXRF (total reflection X ray fluorescence) and XRD (X-ray diffraction). Major elements detected in the precipitates were Al (200 g/kg), S (60 g/kg), and Cu (6 g/kg). Likely solid phases include hydrous amorphous phases of aluminum hydroxides and sulfates, and copper hydroxides/carbonates. Efforts are undergoing to find the optimal mixing ratios between the acidic stream and more alkaline streams to maximize attenuation of dissolved copper. The results of this research could be used for enhancing in-stream natural attenuation of copper and reducing treatment needs at the drinking water facility. Acknowledgements to Fondecyt 1130936 and Conicyt Fondap 15110020

  20. 49 CFR 230.22 - Accident reports.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Requirements § 230.22 Accident reports. In the case of an accident due to failure, from any cause, of a steam locomotive boiler or any part or appurtenance thereof, resulting in serious injury or death to one or...

  1. 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.

  2. Determinants of injuries in passenger vessel accidents.

    PubMed

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

    2015-09-01

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

  3. Determinants of injuries in passenger vessel accidents.

    PubMed

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

    2015-09-01

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

  4. NASA Medical Response to Human Spacecraft Accidents

    NASA Technical Reports Server (NTRS)

    Patlach, Robert

    2010-01-01

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

  5. 48 CFR 836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  6. 48 CFR 636.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Accident prevention. 636... CONTRACTING CONSTRUCTION AND ARCHITECT-ENGINEER CONTRACTS Contract Clauses 636.513 Accident prevention. (a) In... contracting activities shall insert DOSAR 652.236-70, Accident Prevention, in lieu of FAR clause...

  7. 48 CFR 1836.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  8. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

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

  9. 28 CFR 301.106 - Repetitious accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

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

  10. Car Accident Reconstruction and Head Injury Correlation

    NASA Astrophysics Data System (ADS)

    Chawla, A.; Grover, V.; Mukherjee, S.; Hassan, A. M.

    2013-04-01

    Estimation of brain damage remains an elusive issue and controlled tests leading to brain damage cannot be carried out on volunteers. This study reconstructs real-world car accidents to estimate the kinematics of the head impact. This data is to be used to estimate the head injury measures through computer simulations and then correlate reported skull as well as brain damage to impact measures; whence validating the head FE model (Willinger, IJCrash 8:605-617, 2003). In this study, two crash cases were reconstructed. Injury correlation was successful in one of these cases in that the injuries to the brain of one of the car drivers could be correlated in terms of type, location and severity when compared with the tolerance limits of relevant injury parameters (Willinger, IJCrash 8:605-617, 2003).

  11. Estimation Of 137Cs Using Atmospheric Dispersion Models After A Nuclear Reactor Accident

    NASA Astrophysics Data System (ADS)

    Simsek, V.; Kindap, T.; Unal, A.; Pozzoli, L.; Karaca, M.

    2012-04-01

    Nuclear energy will continue to have an important role in the production of electricity in the world as the need of energy grows up. But the safety of power plants will always be a question mark for people because of the accidents happened in the past. Chernobyl nuclear reactor accident which happened in 26 April 1986 was the biggest nuclear accident ever. Because of explosion and fire large quantities of radioactive material was released to the atmosphere. The release of the radioactive particles because of accident affected not only its region but the entire Northern hemisphere. But much of the radioactive material was spread over west USSR and Europe. There are many studies about distribution of radioactive particles and the deposition of radionuclides all over Europe. But this was not true for Turkey especially for the deposition of radionuclides released after Chernobyl nuclear reactor accident and the radiation doses received by people. The aim of this study is to determine the radiation doses received by people living in Turkish territory after Chernobyl nuclear reactor accident and use this method in case of an emergency. For this purpose The Weather Research and Forecasting (WRF) Model was used to simulate meteorological conditions after the accident. The results of WRF which were for the 12 days after accident were used as input data for the HYSPLIT model. NOAA-ARL's (National Oceanic and Atmospheric Administration Air Resources Laboratory) dispersion model HYSPLIT was used to simulate the 137Cs distrubition. The deposition values of 137Cs in our domain after Chernobyl Nuclear Reactor Accident were between 1.2E-37 Bq/m2 and 3.5E+08 Bq/m2. The results showed that Turkey was affected because of the accident especially the Black Sea Region. And the doses were calculated by using GENII-LIN which is multipurpose health physics code.

  12. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis, Model description. Volume 1

    SciTech Connect

    Borkowski, J.A.; Wade, N.L.; Giles, M.M.; Rouhani, S.Z.; Shumway, R.W.; Singer, G.L.; Taylor, D.D.; Weaver, W.L.

    1992-08-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODl version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MODI version produces results consistent with previous versions. Assessment calculations using the two TRAC-BF1 versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  13. TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User`s guide. Volume 2

    SciTech Connect

    Rettig, W.H.; Wade, N.L.

    1992-06-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

  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. 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

  16. 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.

  17. Medical management principles for radiation accidents.

    PubMed

    Meineke, Viktor; van Beuningen, Dirk; Sohns, Torsten; Fliedner, Theodor M

    2003-03-01

    The medical management of radiation accidents requires intensive planning and action. This article looks at the medical management of recent radiation accidents to derive principles for structuring and organizing the treatment of patients who may have radiation-induced health impairments. Although the radiation accidents in Tokai-mura, Japan and Lilo, Georgia were small-scale accidents, they illustrate important and characteristic symptoms and clinical developments. There are lessons to be learned and conclusions to be drawn for the military medical officers concerned with problems of medical management after radiation accidents.

  18. Exploratory analysis of Spanish energetic mining accidents.

    PubMed

    Sanmiquel, Lluís; Freijo, Modesto; Rossell, Josep M

    2012-01-01

    Using data on work accidents and annual mining statistics, the paper studies work-related accidents in the Spanish energetic mining sector in 1999-2008. The following 3 parameters are considered: age, experience and size of the mine (in number of workers) where the accident took place. The main objective of this paper is to show the relationship between different accident indicators: risk index (as an expression of the incidence), average duration index for the age and size of the mine variables (as a measure of the seriousness of an accident), and the gravity index for the various sizes of mines (which measures the seriousness of an accident, too). The conclusions of this study could be useful to develop suitable prevention policies that would contribute towards a decrease in work-related accidents in the Spanish energetic mining industry. PMID:22721539

  19. The Concept of Accident Proneness: A Review

    PubMed Central

    Froggatt, Peter; Smiley, James A.

    1964-01-01

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

  20. Sensitivity Analyses in Small Break LOCA with HPI-Failure: Effect of Break-Size in Secondary-Side Depressurization

    NASA Astrophysics Data System (ADS)

    Kinoshita, Ikuo; Torige, Toshihide; Yamada, Minoru

    2014-06-01

    In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.

  1. Accident/Mishap Investigation System

    NASA Technical Reports Server (NTRS)

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

    2007-01-01

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

  2. 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.

  3. 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.

  4. Temporal Statistic of Traffic Accidents in Turkey

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

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

  5. 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.

  6. 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.

  7. 10 CFR 55.46 - Simulation facilities.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ..., transient, and accident conditions to which the simulator has been designed to respond. The plant-referenced... simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that replicate the...

  8. 10 CFR 55.46 - Simulation facilities.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ..., transient, and accident conditions to which the simulator has been designed to respond. The plant-referenced... simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that replicate the...

  9. 10 CFR 55.46 - Simulation facilities.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ..., transient, and accident conditions to which the simulator has been designed to respond. The plant-referenced... simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that replicate the...

  10. 10 CFR 55.46 - Simulation facilities.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ..., transient, and accident conditions to which the simulator has been designed to respond. The plant-referenced... simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that replicate the...

  11. MSHA releases data on CM crushing accidents

    SciTech Connect

    2007-02-15

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

  12. Cardiac damage presenting late after road accidents.

    PubMed Central

    Mackintosh, A F; Fleming, H A

    1981-01-01

    Six examples of cardiac damage secondary to non-penetrating trauma in road accidents are described. In all six cases the lesion was not recognised at the time of the accident but became clinically important two days to 17 years later. As the patients were young or had unusual lesions, the damage could be attributed to the accident. In older patients with common cardiac problems the trauma might not be recognised as the underlying cause. PMID:7330802

  13. Transportation accident scenarios for commercial spent fuel

    SciTech Connect

    Wilmot, E L

    1981-02-01

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

  14. [Medicolegal and compensation scientific approach to automobile accident].

    PubMed

    Yamanouchi, Haruo

    2002-09-01

    Multiple injuries are frequently observed over the whole body of traffic victims in medico-legal autopsy cases. The assessment of a traffic casualty must include not only the victim but also the vehicle and the circumstances of the accident. Only consideration of all available data permits a better assessment of the mechanism of the crash and causation of injuries. J. M. Thevenet drove the first car carried from France to Japan on February 6, 1898. On October 28th, 1905, the first death by a road traffic accident occurred in Osaka. We performed a retrospective analysis of 279 traffic fatalities examined by medico-legal autopsy in Niigata that occurred over a twenty-two-year period from 1980 to 2001. All persons who had an ICD-10 code were grouped by 153 pedestrians, 43 pedal cyclists, 20 motorcycle riders, 45 car occupants, 11 occupants of pick-up trucks or vans, 4 occupants of heavy transport vehicles and 3 others. The average of ISS (injury scale score) is 40.7 in pedestrians, 26.7 in pedal cyclists, 32.4 in motorcycle riders, 25.1 in car occupants, 16.5 in occupants of pick-up trucks or vans, 24.0 in heavy transport vehicles and 69.0 in others. Rib fractures were observed in 170 cases (60.9%) and the frequency of other injuries was shown in Table 2. Criminal Punishment for drivers involved in 261 traffic accidents amounted to 35 sentences of imprisonment (13.4%), 46 suspension of execution of sentence (17.6%) and 60 sentence of fine (23.0%). Forty prone pedestrians run over by cars showed high ethanol levels in their blood. It was necessary to identify the driver of a vehicle in twelve car accidents and simulation with a computer is very useful. The average of ISS was 34.0 in ten drivers and 22.0 in fourteen fellow passengers. Four sudden natural deaths of drivers at the wheel, eight cases of death immediately after and from one day to five months after road traffic accidents, nine suicides and one intentional accident are excluded from traffic death. Both a

  15. [Medicolegal and compensation scientific approach to automobile accident].

    PubMed

    Yamanouchi, Haruo

    2002-09-01

    Multiple injuries are frequently observed over the whole body of traffic victims in medico-legal autopsy cases. The assessment of a traffic casualty must include not only the victim but also the vehicle and the circumstances of the accident. Only consideration of all available data permits a better assessment of the mechanism of the crash and causation of injuries. J. M. Thevenet drove the first car carried from France to Japan on February 6, 1898. On October 28th, 1905, the first death by a road traffic accident occurred in Osaka. We performed a retrospective analysis of 279 traffic fatalities examined by medico-legal autopsy in Niigata that occurred over a twenty-two-year period from 1980 to 2001. All persons who had an ICD-10 code were grouped by 153 pedestrians, 43 pedal cyclists, 20 motorcycle riders, 45 car occupants, 11 occupants of pick-up trucks or vans, 4 occupants of heavy transport vehicles and 3 others. The average of ISS (injury scale score) is 40.7 in pedestrians, 26.7 in pedal cyclists, 32.4 in motorcycle riders, 25.1 in car occupants, 16.5 in occupants of pick-up trucks or vans, 24.0 in heavy transport vehicles and 69.0 in others. Rib fractures were observed in 170 cases (60.9%) and the frequency of other injuries was shown in Table 2. Criminal Punishment for drivers involved in 261 traffic accidents amounted to 35 sentences of imprisonment (13.4%), 46 suspension of execution of sentence (17.6%) and 60 sentence of fine (23.0%). Forty prone pedestrians run over by cars showed high ethanol levels in their blood. It was necessary to identify the driver of a vehicle in twelve car accidents and simulation with a computer is very useful. The average of ISS was 34.0 in ten drivers and 22.0 in fourteen fellow passengers. Four sudden natural deaths of drivers at the wheel, eight cases of death immediately after and from one day to five months after road traffic accidents, nine suicides and one intentional accident are excluded from traffic death. Both a

  16. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  17. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

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

    2015-04-16

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

  18. 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.

  19. Industrial Safety and Accidents Prevention

    SciTech Connect

    Sajjad Akbar

    2006-07-01

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

  20. HTGR severe accident sequence analysis

    SciTech Connect

    Harrington, R.M.; Ball, S.J.; Kornegay, F.C.

    1982-01-01

    Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.

  1. The determinants of fishing vessel accident severity.

    PubMed

    Jin, Di

    2014-05-01

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

  2. Thermal hydraulic features of the TMI accident

    NASA Astrophysics Data System (ADS)

    Tolman, B.

    1985-10-01

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

  3. 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.

  4. [SAFETY IN THE ELDERLY: HOME ACCIDENTS].

    PubMed

    Martín-Espinosa, Noelia M; Píriz-Campos, Rosa Ma; Cordeiro, Raú; Muñoz Bermejo, Laura; Casado Verdjo, Inés; Postigo Mota, Salvador

    2016-05-01

    Home accidents are more common in the elderly and they can have serious consequences to the injured person's health. At home, chances to suffer accidents of any type are higher, because it's the place where old people spend most of their daily time. It is important to point out that a high percentage of domestic accidents could be easily avoided by taking some simple cautions. The main aim of this paper is to know how we can prevent most common domestic accidents in the aged population: falls, burnings, poisonings and fire prevention. PMID:27405149

  5. The determinants of fishing vessel accident severity.

    PubMed

    Jin, Di

    2014-05-01

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

  6. 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.

  7. Biorhythmic Cycles and the Incidence of Industrial Accidents

    ERIC Educational Resources Information Center

    Carvey, Davis W.; Nibler, Roger G.

    1977-01-01

    The biorhythm theory of accident explanation that has been increasingly popularized in the business press was empirically examined. Municipal employees involved in work-related vehicular accidents and in on-the-job accidents provided the data. Each accident was analyzed to determine whether or not the accident occurred on a biorhythmically…

  8. Predicted spatio-temporal dynamics of radiocesium deposited onto forests following the Fukushima nuclear accident

    PubMed Central

    Hashimoto, Shoji; Matsuura, Toshiya; Nanko, Kazuki; Linkov, Igor; Shaw, George; Kaneko, Shinji

    2013-01-01

    The majority of the area contaminated by the Fukushima Dai-ichi nuclear power plant accident is covered by forest. To facilitate effective countermeasure strategies to mitigate forest contamination, we simulated the spatio-temporal dynamics of radiocesium deposited into Japanese forest ecosystems in 2011 using a model that was developed after the Chernobyl accident in 1986. The simulation revealed that the radiocesium inventories in tree and soil surface organic layer components drop rapidly during the first two years after the fallout. Over a period of one to two years, the radiocesium is predicted to move from the tree and surface organic soil to the mineral soil, which eventually becomes the largest radiocesium reservoir within forest ecosystems. Although the uncertainty of our simulations should be considered, the results provide a basis for understanding and anticipating the future dynamics of radiocesium in Japanese forests following the Fukushima accident. PMID:23995073

  9. 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.

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

    SciTech Connect

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

    2003-01-01

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

  11. Assessment of the risk due to release of carbon fiber in civil aircraft accidents, phase 2

    NASA Technical Reports Server (NTRS)

    Pocinki, L.; Cornell, M. E.; Kaplan, L.

    1980-01-01

    The risk associated with the potential use of carbon fiber composite material in commercial jet aircraft is investigated. A simulation model developed to generate risk profiles for several airports is described. The risk profiles show the probability that the cost due to accidents in any year exceeds a given amount. The computer model simulates aircraft accidents with fire, release of fibers, their downwind transport and infiltration of buildings, equipment failures, and resulting ecomomic impact. The individual airport results were combined to yield the national risk profile.

  12. A Serious Game for Traffic Accident Investigators

    ERIC Educational Resources Information Center

    Binsubaih, Ahmed; Maddock, Steve; Romano, Daniela

    2006-01-01

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

  13. Normal Accident at Three Mile Island.

    ERIC Educational Resources Information Center

    Perrow, Charles

    1981-01-01

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

  14. 48 CFR 36.513 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  15. Global estimates of fatal occupational accidents.

    PubMed

    Takala, J

    1999-09-01

    Data on occupational accidents are not available from all countries in the world. Furthermore, underreporting, limited coverage by reporting and compensation schemes, and non-harmonized accident recording and notification systems undermine efforts to obtain worldwide information on occupational accidents. This paper presents a method and new estimated global figures of fatal accidents at work by region. The fatal occupational accident rates reported to the International Labour Office are extended to the total employed workforce in countries and regions. For areas not covered by the reported information, rates from other countries that have similar or comparable conditions are applied. In 1994, an average estimated fatal occupational accident rate in the whole world was 14.0 per 100,000 workers, and the total estimated number of fatal occupational accidents was 335,000. The rates are different for individual countries and regions and for separate branches of economic activity. In conclusion, fatal occupational accident figures are higher than previously estimated. The new estimates can be gradually improved by obtaining and adding data from countries where information is not yet available. Sectoral estimates for at least key economic branches in individual countries would further increase the accuracy.

  16. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

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

  17. 22 CFR 102.8 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

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

  18. Aircraft accidents.method of analysis

    NASA Technical Reports Server (NTRS)

    1937-01-01

    This report is a revision of NACA-TR-357. It was prepared by the Committee on Aircraft Accidents. The purpose of this report is to provide a basis for the classification and comparison of aircraft accidents, both civil and military.

  19. Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

    SciTech Connect

    Kevin G. Field; Richard Howard; Michael Teague

    2014-06-01

    The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items to form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.

  20. Calculation of total effective dose equivalent and collective dose in the event of a LOCA in Bushehr Nuclear Power Plant.

    PubMed

    Raisali, G; Davilu, H; Haghighishad, A; Khodadadi, R; Sabet, M

    2006-01-01

    In this research, total effective dose equivalent (TEDE) and collective dose (CD) are calculated for the most adverse potential accident in Bushehr Nuclear Power Plant from the viewpoint of radionuclides release to the environment. Calculations are performed using a Gaussian diffusion model and a slightly modified version of AIREM computer code to adopt for conditions in Bushehr. The results are comparable with the final safety analysis report which used DOZAM code. Results of our calculations show no excessive dose in populated regions. Maximum TEDE is determined to be in the WSW direction. CD in the area around the nuclear power plant by a distance of 30 km (138 man Sv) is far below the accepted limits. Thyroid equivalent dose is also calculated for the WSW direction (maximum 25.6 mSv) and is below the limits at various distances from the reactor stack.

  1. Commercial SNF Accident Release Fractions

    SciTech Connect

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the

  2. Road accidents caused by drivers falling asleep.

    PubMed

    Sagberg, F

    1999-11-01

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

  3. [Poisoning accidents with household chemicals among children].

    PubMed

    Johannsen, H G; Mikkelsen, J B

    1994-10-01

    A review is presented of the registration of all poisoning accidents among children aged 0-6 years treated at the University Hospital, Odense, Denmark during the period 1.1.1980-31.12.1992. There were 1751 poisoning accidents of which 482 were accidents with household chemicals. There were 69 accidents with lamp oil (Petroleum) of which 67 were in the age group 0-3 years. A peak incidence in the age group 0-3 years old is seen in 1986. In 1992 the incidence is at about the same level as in 1980. The incidence in the age group 4-6 years is at almost the same level throughout the entire period. We conclude that it is necessary to continue with campaigns to prevent accidents with household chemicals among children.

  4. [Clinical examinations for the traffic accident patients].

    PubMed

    Hitosugi, Masahito

    2008-11-30

    Traffic accident is a leading cause of unintentional death and about six-thousands annually died in Japan. As about one-million of persons suffer from traffic injuries, most of them seek medical attention. Therefore, medical staffs have to find the injuries accurately and treat immediately. Furthermore, the cause of accident should also be considered; why the accident was occurred, human error of the driver? To solve these problems, clinical examinations were needed. Medical staffs have to understand the characteristics of the traffic injuries: severe and multiple blunt injuries, popular injuries can be estimated with considering the pattern of the accident. Because some of the accidents are occurred when the driver is under the influence of alcohol and other drugs, screening of these subjects should be performed. Because the public is largely unaware of the preventable nature of traffic injuries, in addition to diagnose and treat accurately, we medical staffs have to attend on the primary prevention of the traffic injuries.

  5. Accident prevention: the health visitor's role.

    PubMed

    Levene, S

    1992-10-01

    The health of the nation white paper sets targets in five key areas for reductions in both mortality and morbidity: coronary heart disease and stroke, cancers, mental illness, HIV/Aids and sexual health and accidents. In a series of articles in Health visitor, experts will be considering the opportunities the white paper offers for community nurses in each of the key areas. Here Dr Sara Levene, medical consultant to the Child Accident Prevention Trust, considers accidents, a major problem which health visitors can do much to control. She reviews how accidents are presented in the white paper, what health visitors can do and what resources are available to help them. She offers particular advice on special accident prevention initiatives and discusses some of the opportunities created by the white paper.

  6. Pilot-error accidents: male vs female.

    PubMed

    Vail, G J; Ekman, L G

    1986-12-01

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

  7. Human Factors in Cabin Accident Investigations

    NASA Technical Reports Server (NTRS)

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

    1996-01-01

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

  8. [Cerebral vascular accidents in French Polynesia].

    PubMed

    Gras, C; Papouin, G; Prigent, D; Beaugendre, E; Lionet, P; Brodin, S; Legall, R; Marjou, F; Spiegel, A; Gendron, Y

    1992-01-01

    The authors report on the results of a survey on cardiovascular accidents hospitalized between 01 April 1990 and 31 January 1991 carried out in the Services of Medicine and Cardiology in the Territorial Hospital Center of Papeete. This survey was: 56 cardiovascular accidents: 1/4 (hemorrhagic and 3/4 (42) ischemic. Mean age 59 (extremes 23-86). 36 males (64%); 20 females (36%). 50 Polynesians; 6 Chinese people. Among the risk factors recorded, 38 (68%) were hypertensed patients; 17 (30%) were due to tabagism and 15 (25%) to diabetes; 3 (5%) are known to be carriers of a hypercholesterolemia. 59% of the patients had no case history; 25% the cardiovascular accidents have been observed in patients with cardiopathy; 12.5% are recurrent cardiovascular accidents. Clinically, 5 transient ischemic accidents (12%) out of 42 cardiovascular ischemic accidents. High arterial tension was recognized in 12/14 (86%) of hemorrhagic cardiovascular accidents and in 26/42 (62%) of ischemic cardiovascular accidents. In 42 ischemic cardiovascular accidents, 31 patients suffered from cardiopathy (74%) of which 15 (36%) presented an embolic cardiopathy. Interest of echography and electrocardiogram are discussed. Ultrasonic exam of carotid vessels was found abnormal in almost half of the cases when utilized (12/26). Finally, etiological diagnosis was certain in 17 cases, of presumption in 16 cases, and in 9 cases, it was not possible to precise any cardiovascular etiology. Tomodensitometric tests are discussed. 86% of the ischemic cardiovascular accident were treated with anticoagulants/thrombocyte antiagglutination. 24% of the patients died, 50% recovered incompletely and 26% completely. PMID:1602953

  9. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect

    Rollstin, J.A. ); Chanin, D.I. ); Jow, H.N. )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management.

  10. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect

    Chanin, D.I. ); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  11. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. ); Rollstin, J.A. ); Chanin, D.I. )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  12. Code System for Toxic Gas Accident Analysis.

    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

  13. The Fukushima accident was preventable.

    PubMed

    Synolakis, Costas; Kânoğlu, Utku

    2015-10-28

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

  14. The Fukushima accident was preventable.

    PubMed

    Synolakis, Costas; Kânoğlu, Utku

    2015-10-28

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

  15. [A study on fall accident].

    PubMed

    Lee, H S; Kim, M J

    1997-01-01

    The study was conducted from November 1995 to May 1996 at the one general hospital in Seoul. The total subjects of this study were 412 patients who have the experience of fall accident, among them 31 was who have fallen during hospitalization and 381 was who visited emergency room and out patient clinic. The purposes of this study were to determine the characteristics, risk factors and results of fall accident and to suggest the nursing strategies for prevention of fall. Data were collected by reviewing the medical records and interviewing with the fallers and their family members. For data analysis spss/pc+ program was utilized for descriptive statistics, adjusted standardized X2-test. The results of this study were as follows: 1) Total subjects were 412 fallers, of which 245 (59.5%) were men and 167 (40.5%) were women. Age were 0-14 years 79 (19.2%), 15-44 years 125 (30.4%), 45-64 years 104 (25.2%), over 65 years 104 (25.2%). 2) There was significant association between age and the sexes (X2 = 39.17, P = 0.00). 3) There was significant association between age and history of falls (X2 = 44.41, P = .00). And history of falls in the elderly was significantly associated with falls. 4) There was significant association with age and medical diagnosis (X2 = 140.66, P = .00), chief medical diagnosis were hypertension (34), diabetes mellitus (22), arthritis (11), stroke (8), fracture (7), pulmonary tuberculosis (6), dementia (5) and cataract (5). 5) There was significant association between age and intrinsic factors: cognitive impairment, mobility impairment, insomnia, emotional problems, urinary difficulty, visual impairments, hearing impairments, use of drugs (sedatives, antihypertensive drugs, diuretics, antidepressants) (P < 0.05). But there was no significant association between age and dizziness (X2 = 2.87, P = .41). 6) 15.3% of total fallers were drunken state when they were fallen. 7) Environmental factors of fall accident were unusual posture (50.9%), slips (35

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

    SciTech Connect

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

    1986-06-01

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

  17. 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.

  18. Accidents in Canada: mortality and hospitalization.

    PubMed

    Riley, R; Paddon, P

    1989-01-01

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

  19. Road accidents and business cycles in Spain.

    PubMed

    Rodríguez-López, Jesús; Marrero, Gustavo A; González, Rosa Marina; Leal-Linares, Teresa

    2016-11-01

    This paper explores the causes behind the downturn in road accidents in Spain across the last decade. Possible causes are grouped into three categories: Institutional factors (a Penalty Point System, PPS, dating from 2006), technological factors (active safety and passive safety of vehicles), and macroeconomic factors (the Great recession starting in 2008, and an increase in fuel prices during the spring of 2008). The PPS has been blessed by incumbent authorities as responsible for the decline of road fatalities in Spain. Using cointegration techniques, the GDP growth rate, the fuel price, the PPS, and technological items embedded in motor vehicles appear to be statistically significantly related with accidents. Importantly, PPS is found to be significant in reducing fatal accidents. However, PPS is not significant for non-fatal accidents. In view of these results, we conclude that road accidents in Spain are very sensitive to the business cycle, and that the PPS influenced the severity (fatality) rather than the quantity of accidents in Spain. Importantly, technological items help explain a sizable fraction in accidents downturn, their effects dating back from the end of the nineties.

  20. Childhood accidents: epidemiology, trends, and prevention.

    PubMed Central

    Kemp, A; Sibert, J

    1997-01-01

    Accidents are the most common cause of death in children over one year of age. Prevention remains a high priority. We have reviewed the current epidemiology of childhood accidents and their prevention, and made recommendations for the future. In 1992, 559 children died in United Kingdom as a result of an accidents--240 from road traffic accidents and 100 from burns and scalds. Every year 50 children drown. Accidents cause significant disability to children. Many children, up to one in four of the population in urban areas, attend accident and emergency departments, and 5-10% of these are admitted to hospital. Accident risk factors include low social class, psychosocial stress, an unsafe environment, and child developmental disorders. Research has shown that prevention is best achieved by making the child's environment safer, often through legislation. Insufficient resources have been put into both research into childhood injuries and preventive work in communities. Collaboration between health authorities, NHS trusts, local authorities and community networks is vital if success is to be achieved. A national safety agenda for children would focus the attention that this problem deserves. PMID:9315935

  1. Accidents in Canada: mortality and hospitalization.

    PubMed

    Riley, R; Paddon, P

    1989-01-01

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

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

    SciTech Connect

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

    1984-01-01

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

  3. Bayes classifiers for imbalanced traffic accidents datasets.

    PubMed

    Mujalli, Randa Oqab; López, Griselda; Garach, Laura

    2016-03-01

    Traffic accidents data sets are usually imbalanced, where the number of instances classified under the killed or severe injuries class (minority) is much lower than those classified under the slight injuries class (majority). This, however, supposes a challenging problem for classification algorithms and may cause obtaining a model that well cover the slight injuries instances whereas the killed or severe injuries instances are misclassified frequently. Based on traffic accidents data collected on urban and suburban roads in Jordan for three years (2009-2011); three different data balancing techniques were used: under-sampling which removes some instances of the majority class, oversampling which creates new instances of the minority class and a mix technique that combines both. In addition, different Bayes classifiers were compared for the different imbalanced and balanced data sets: Averaged One-Dependence Estimators, Weightily Average One-Dependence Estimators, and Bayesian networks in order to identify factors that affect the severity of an accident. The results indicated that using the balanced data sets, especially those created using oversampling techniques, with Bayesian networks improved classifying a traffic accident according to its severity and reduced the misclassification of killed and severe injuries instances. On the other hand, the following variables were found to contribute to the occurrence of a killed causality or a severe injury in a traffic accident: number of vehicles involved, accident pattern, number of directions, accident type, lighting, surface condition, and speed limit. This work, to the knowledge of the authors, is the first that aims at analyzing historical data records for traffic accidents occurring in Jordan and the first to apply balancing techniques to analyze injury severity of traffic accidents.

  4. Chemical considerations in severe accident analysis

    SciTech Connect

    Malinauskas, A.P.; Kress, T.S.

    1988-01-01

    The Reactor Safety Study presented the first systematic attempt to include fission product physicochemical effects in the determination of expected consequences of hypothetical nuclear reactor power plant accidents. At the time, however, the data base was sparse, and the treatment of fission product behavior was not entirely consistent or accurate. Considerable research has since been performed to identify and understand chemical phenomena that can occur in the course of a nuclear reactor accident, and how these phenomena affect fission product behavior. In this report, the current status of our understanding of the chemistry of fission products in severe core damage accidents is summarized and contrasted with that of the Reactor Safety Study.

  5. Impact of nuclear accidents on marine biota.

    PubMed

    Vives i Batlle, Jordi

    2011-07-01

    The accident at the Fukushima Daiichi nuclear power plant, precipitated by the earthquake and subsequent tsunami that struck the northeastern coast of Japan in March 2011, has raised concerns about the potential impact to marine biota posed by the release of radioactive water and radionuclide particles into the environment. The Fukushima accident is the only major nuclear accident that has resulted in the direct discharge of radioactive materials into a coastal environment. This article briefly summarizes what is currently understood about the effects of radioactive wastewaters and radionuclides to marine life.

  6. Post Quench Ductility Evaluation of Zircaloy-4 and Select Iron Alloys under Design Basis and Extended LOCA Conditions

    SciTech Connect

    Yan, Yong; Keiser, James R; Terrani, Kurt A; Bell, Gary L; Snead, Lance

    2014-01-01

    Oxidation experiments were conducted at 1200 C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 were determined to be k = 2.173 107 g2/cm4/s C, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post quenched samples was evaluated by ring compression tests at 135 C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.

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

    NASA Technical Reports Server (NTRS)

    Evans, Joni K.

    2007-01-01

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

  8. 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.

  9. Weather and Dispersion Modeling of the Fukushima Daiichi Nuclear Power Station Accident

    NASA Astrophysics Data System (ADS)

    Dunn, Thomas; Businger, Steven

    2014-05-01

    The surface deposition of radioactive material from the accident at the Fukushima Daiichi nuclear power station was investigated for 11 March to 17 March 2011. A coupled weather and dispersion modeling system was developed and simulations of the accident performed using two independent source terms that differed in emission rate and height and in the total amount of radioactive material released. Observations in Japan during the first week of the accident revealed a natural grouping between periods of dry (12-14 March) and wet (15-17 March) weather. The distinct weather regimes served as convenient validation periods for the model predictions. Results show significant differences in the distribution of cumulative surface deposition of 137Cs due to wet and dry removal processes. A comparison of 137Cs deposition predicted by the model with aircraft observations of surface-deposited gamma radiation showed reasonable agreement in surface contamination patterns during the dry phase of the accident for both source terms. It is suggested that this agreement is because of the weather model's ability to simulate the extent and timing of onshore flow associated with a sea breeze circulation that developed around the time of the first reactor explosion. During the wet phase of the accident the pattern is not as well predicted. It is suggested that this discrepancy is because of differences between model predicted and observed precipitation distributions.

  10. Traffic accidents on a single-lane road with multi-slowdown sections

    NASA Astrophysics Data System (ADS)

    Li, Xingli; Kuang, Hua; Fan, Yanhong; Zhang, Guoxin

    2014-02-01

    In this paper, an extended cellular automaton model is proposed to simulate the complex characteristics of traffic flow and the probability of the occurrence of traffic accidents by considering the modified conditions for determining whether traffic accidents happen and the effect of multi-slowdown sections on a highway. The simulation results show that the multi-slowdown sections can lead to multiphase coexistences (i.e. free flow phase, congestion phase and saturation phase) in traffic system. The fundamental diagram shows that the number of slowdown section does not influence the mean velocity and the mean flow under the periodic boundary condition, but the existence of slowdown sections can effectively reduce the occurrence of traffic accident. In particular, it is found that the probability of car accidents to occur is the largest at the joint of the normal-speed section and slowdown section, and the underlying mechanism is analyzed. In addition, to design the appropriate limited speed and reduce the differences between the normal speed and limited speed will alleviate traffic congestion and reduce the occurrence of traffic accidents obviously.

  11. Review of models applicable to accident aerosols

    SciTech Connect

    Glissmeyer, J.A.

    1983-07-01

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

  12. 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.

  13. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

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

    1999-11-01

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

  14. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect

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

    1999-01-01

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

  15. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

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

  16. 50 CFR 25.72 - Reporting of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  17. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

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

  18. 40 CFR 68.42 - Five-year accident history.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 16 2014-07-01 2014-07-01 false Five-year accident history. 68.42... (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.42 Five-year accident history. (a) The owner or operator shall include in the five-year accident history all accidental releases...

  19. 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…

  20. 48 CFR 852.236-87 - Accident prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

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

    Code of Federal Regulations, 2010 CFR

    2010-07-01

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

  2. 46 CFR 97.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  3. 46 CFR 196.30-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  4. 46 CFR 78.33-5 - Accidents to machinery.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

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

  5. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 4 2012-01-01 2012-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  6. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  7. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 4 2013-01-01 2013-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  8. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 4 2014-01-01 2014-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... nuclear accident is possible, shall provide nuclear accident dosimetry for those individuals. (b)...

  9. 10 CFR 835.1304 - Nuclear accident dosimetry.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

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

    NASA Astrophysics Data System (ADS)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

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

  11. Environmental measurements during the TMI-2 accident

    SciTech Connect

    Hull, A.P.

    1988-01-01

    Although the environmental consequences of the TMI accident were relatively insignificant, it was a major test of the ability of the involved state and federal radiological agencies to make a coordinated environmental monitoring response. This was accomplished largely on an ad hoc basis under the leadership of DOE. With some fine tuning, it is the basis for today's integrated FRMAP monitoring plan, which would be put into operation should another major accident occur at a US nuclear facility.

  12. [Dysbaric accident in deep sea fishing].

    PubMed

    López Oblaré, B; Campos Pascual, F

    1995-05-20

    The case of a dysbaric accident with occurred in a professional athlete during a national competition is herein reported. The clinical symptoms and response to treatment in a depressurization chamber in addition to CT controls should alert physicians in coastal areas in which this sport is carried out in order to take into consideration neurologic disorders which may be due to dysbaric accidents such as those which occur in scuba divers.

  13. MELCOR analyses for accident progression issues

    SciTech Connect

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

    1991-01-01

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

  14. Truck accident involving unirradiated nuclear fuel

    SciTech Connect

    Carlson, R.W.; Fischer, L.E.

    1992-07-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 nuclear fuel assemblies in 12 containers on Interstate 1-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the.tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1300{degrees}F and 1800{degrees}F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk.

  15. Truck accident involving unirradiated nuclear fuel

    SciTech Connect

    Carlson, R.W.; Fischer, L.E.

    1992-07-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 nuclear fuel assemblies in 12 containers on Interstate 1-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the.tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1300[degrees]F and 1800[degrees]F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk.

  16. 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

  17. 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.

  18. Multibody model of the human upper extremity for fracture simulation

    PubMed Central

    Milanowicz, Marcin; Kędzior, Krzysztof

    2016-01-01

    About 3.8 million people are injured in accidents at work in Europe every year. The resulting high costs are incurred by the victims themselves, their families, employers and society. We have used a numerical simulation to reconstruct accidents at work for several years. To reconstruct these accidents MADYMO R7.5 with a numerical human model (pedestrian model) is used. However, this model is dedicated to the analysis of car-to-pedestrian accidents and thus cannot be fully used for reconstructing accidents at work. Therefore, we started working on the development of a numerical model of the human body for the purpose of simulating accidents at work. Developing a new numerical model which gives an opportunity to simulate fractures of the upper extremity bones is a stage of that work. PMID:26651896

  19. Lockout/tagout accident investigation.

    PubMed

    White, James R

    2014-08-01

    When I was in boot camp, our drill instructor told us that assume makes an ass out of u and me. It was true then, and it is true today. In this instance, assumptions came into play several times, both by the worker and by the companies involved. The good news is that it did not result in a fatality, but that does not relieve the pain and suffering that the employee had to endure. This same type of scenario is likely repeated at many job sites throughout the United States. Multiple contractors, dozens--maybe hundreds--of workers, power system equipment and devices; all of these have to be taken into consideration when performing maintenance activities. It can become a blur. People are people, and people make mistakes. That is why we have OSHA regulations, NFPA 70E, company procedures, policies, etc. Most if not all of us have either been involved in accidents or know people who have been. It's not like it's a secret that people make mistakes, but talk to some and they seem to think only others have that failing. Safety is not about just any one procedure or rule. It's about slowing down, making a plan, and executing that plan. There are plenty of tools available to help us: policies, procedures, codes, standards, federal regulations, and state and local laws. I am not about to say that the worker involved in this incident was not taking safety seriously, but he failed to follow some fundamental safety rules like test-before-touch. If he had taken just that one step, there would be nothing to write about. PMID:25188988

  20. Drudgery, accidents and injuries in Indian agriculture.

    PubMed

    Nag, Pranab Kumar; Nag, Anjali

    2004-04-01

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

  1. [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

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

    PubMed

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

    1990-03-01

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

  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. Long-term transport and dispersion of 137Cs released into ocean off Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Zhao, Chang; Qiao, Fangli; Wang, Guansuo; Xia, Changshui; Jung, KyungTae

    2014-05-01

    In the following days after the Fukushima nuclear accident which happened in 11th March 2011, significant amounts of radioactive materials (131I, 134Cs and 137Cs) had been leaking into the terrestrial and marine environments. The radionuclides model was used to study the distribution of the 137Cs in the Pacific and the Indian Ocean released from the Fukushima accident. The simulation on the distribution of 137Cs agrees well with the the observed profiles in the 9th November 2011, which proved the validaty of the model. In the first year of our model run, the 137Cs is carried eastward by the Kuroshio and its extension, spreading southward and northword meanwhile. Four or five years after the accident, the 137Cs reaches the US coast with the surface waters of the Pacific Ocean; its concentration is no higher than 3 Bq/m3. Ten years after the accident, all the North Pacific Ocean is labeled with the 137Cs from the Fukushima. The concentration is less than 1 Bq/m3 at that time. Thirty years after the accident, the concentration of 137Cs in both the Pacific and the Indian Ocean is below 0.1 Bq/m3. Since the spreading path of 137Cs from the Fukushima nuclear accident is just the migration route of the Pacific tuna, a kind of fish inhabit the western and eastern North Pacific, it may cause radioactive contamination to the fish. In the offshore seas of China, the 137Cs from Fukushima nuclear accident is very low (<0.2 Bq/m3) .

  5. 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.

  6. Emission of spherical cesium-bearing particles from an early stage of the Fukushima nuclear accident

    PubMed Central

    Adachi, Kouji; Kajino, Mizuo; Zaizen, Yuji; Igarashi, Yasuhito

    2013-01-01

    The Fukushima nuclear accident released radioactive materials into the environment over the entire Northern Hemisphere in March 2011, and the Japanese government is spending large amounts of money to clean up the contaminated residential areas and agricultural fields. However, we still do not know the exact physical and chemical properties of the radioactive materials. This study directly observed spherical Cs-bearing particles emitted during a relatively early stage (March 14–15) of the accident. In contrast to the Cs-bearing radioactive materials that are currently assumed, these particles are larger, contain Fe, Zn, and Cs, and are water insoluble. Our simulation indicates that the spherical Cs-bearing particles mainly fell onto the ground by dry deposition. The finding of the spherical Cs particles will be a key to understand the processes of the accident and to accurately evaluate the health impacts and the residence time in the environment. PMID:23989894

  7. Twenty-second ORNL intercomparison of criticality accident dosimetry systems, August 12-16, 1985

    SciTech Connect

    Swaja, R.E.; Oyan, R.; Sims, C.S.

    1986-05-01

    The twenty-second in a series of criticality accident dosimetry intercomparison studies was conducted at the Oak Ridge National Laboratory's Dosimetry Applications Research Facility during August 12-16, 1985. The Health Physics Research Reactor operated in the pulse mode over Storage Pit No. 1 was used to simulate three criticality accidents with different radiation fields. Participants from nine organizations measured neutron doses between 0.36 and 3.78 Gy and gamma doses between 0.22 and 0.80 Gy at area monitoring stations and on phantoms. Approximately 68% of all neutron dose estimates based on foil activation, thermoluminescent, hair activation, and blood sodium activation methods were within +-25% of reference values. About 44% of all gamma results measured using thermoluminescent dosimeters (TLD-700 or CaSO/sub 4/ phosphors) were within 20% of reference doses. The generally poor measurement accuracy exhibited in this study indicates a need for continuing ORNL accident dosimetry intercomparison and training programs.

  8. Investigating accident causation through information network modelling.

    PubMed

    Griffin, T G C; Young, M S; Stanton, N A

    2010-02-01

    Management of risk in complex domains such as aviation relies heavily on post-event investigations, requiring complex approaches to fully understand the integration of multi-causal, multi-agent and multi-linear accident sequences. The Event Analysis of Systemic Teamwork methodology (EAST; Stanton et al. 2008) offers such an approach based on network models. In this paper, we apply EAST to a well-known aviation accident case study, highlighting communication between agents as a central theme and investigating the potential for finding agents who were key to the accident. Ultimately, this work aims to develop a new model based on distributed situation awareness (DSA) to demonstrate that the risk inherent in a complex system is dependent on the information flowing within it. By identifying key agents and information elements, we can propose proactive design strategies to optimize the flow of information and help work towards avoiding aviation accidents. Statement of Relevance: This paper introduces a novel application of an holistic methodology for understanding aviation accidents. Furthermore, it introduces an ongoing project developing a nonlinear and prospective method that centralises distributed situation awareness and communication as themes. The relevance of findings are discussed in the context of current ergonomic and aviation issues of design, training and human-system interaction. PMID:20099174

  9. The three essentials for accident prevention.

    PubMed

    Eastman, Crystal

    2014-11-01

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

  10. NASA Medical Response to Human Spacecraft Accidents

    NASA Technical Reports Server (NTRS)

    Patlach, Robert

    2011-01-01

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

  11. Single pilot IFR accident data analysis

    NASA Technical Reports Server (NTRS)

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

    1982-01-01

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

  12. 1983 international intercomparison of nuclear accident dosimetry systems at Oak Ridge National Laboratory

    SciTech Connect

    Swaja, R.E.; Greene, R.T.; Sims, C.S.

    1985-04-01

    An international intercomparison of nuclear accident dosimetry systems was conducted during September 12-16, 1983, at Oak Ridge National Laboratory (ORNL) using the Health Physics Research Reactor operated in the pulse mode to simulate criticality accidents. This study marked the twentieth in a series of annual accident dosimetry intercomparisons conducted at ORNL. Participants from ten organizations attended this intercomparison and measured neutron and gamma doses at area monitoring stations and on phantoms for three different shield conditions. Results of this study indicate that foil activation techniques are the most popular and accurate method of determining accident-level neutron doses at area monitoring stations. For personnel monitoring, foil activation, blood sodium activation, and thermoluminescent (TL) methods are all capable of providing accurate dose estimates in a variety of radiation fields. All participants in this study used TLD's to determine gamma doses with very good results on the average. Chemical dosemeters were also shown to be capable of yielding accurate estimates of total neutron plus gamma doses in a variety of radiation fields. While 83% of all neutron measurements satisfied regulatory standards relative to reference values, only 39% of all gamma results satisfied corresponding guidelines for gamma measurements. These results indicate that continued improvement in accident dosimetry evaluation and measurement techniques is needed.

  13. 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.

  14. Agricultural implications of the Fukushima nuclear accident.

    PubMed

    Nakanishi, Tomoko M

    2016-08-01

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

  15. A Review of Criticality Accidents 2000 Revision

    SciTech Connect

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

    2000-05-01

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

  16. Assessment of CRBR core disruptive accident energetics

    SciTech Connect

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

    1984-03-01

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

  17. [Psychosocial aspects and accidents in land transport].

    PubMed

    Morales-Soto, Nelson; Alfaro-Basso, Daniel; Gálvez-Rivero, Wilfredo

    2010-06-01

    Road traffic accidents are a public health problem in Peru, having caused 35 596 deaths in Peru between 1998 and 2008. Lima is the most affected region, presenting 61.7% of the accidents, the annual cost reached one thousand million dollars, equivalent to a third part of the investment in health. Available studies give emphasis to the protagonists--the drivers, the pedestrians--or to equipment and roads; the laws have been modified and containment plans for accidents have been implemented, but the incidence remains the same. We raise the possibility of exploring behavioral and social factors that could be relevant in the genesis of the problem, revising those related to current disorder in transport, the behaviors of drivers and pedestrians and the permissiveness of society in general particularly of the authority. We propose research and a multidisciplinary and intersectoral intervention. PMID:21072481

  18. Structural aspects of the Chernobyl accident

    SciTech Connect

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

    1988-09-02

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

  19. Reconfigurable mobile manipulation for accident response

    SciTech Connect

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

    2000-06-06

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

  20. Allometric scaling and accidents at work

    PubMed Central

    Cempel, Czesław; Tabaszewski, Maciej; Ordysiński, Szymon

    2016-01-01

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

  1. Bundled automobile insurance coverage and accidents.

    PubMed

    Li, Chu-Shiu; Liu, Chwen-Chi; Peng, Sheng-Chang

    2013-01-01

    This paper investigates the characteristics of automobile accidents by taking into account two types of automobile insurance coverage: comprehensive vehicle physical damage insurance and voluntary third-party liability insurance. By using a unique data set in the Taiwanese automobile insurance market, we explore the bundled automobile insurance coverage and the occurrence of claims. It is shown that vehicle physical damage insurance is the major automobile coverage and affects the decision to purchase voluntary liability insurance coverage as a complement. Moreover, policyholders with high vehicle physical damage insurance coverage have a significantly higher probability of filing vehicle damage claims, and if they additionally purchase low voluntary liability insurance coverage, their accident claims probability is higher than those who purchase high voluntary liability insurance coverage. Our empirical results reveal that additional automobile insurance coverage information can capture more driver characteristics and driving behaviors to provide useful information for insurers' underwriting policies and to help analyze the occurrence of automobile accidents.

  2. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect

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

    2013-09-01

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

  3. Agricultural implications of the Fukushima nuclear accident

    PubMed Central

    Nakanishi, Tomoko M.

    2016-01-01

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

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

    SciTech Connect

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

    1995-12-31

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

  5. 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.

  6. 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

  7. Simulating nanoscale semiconductor devices.

    SciTech Connect

    Salinger, Andrew Gerhard; Zhao, P.; Woolard, D. L.; Kelley, C. Tim; Lasater, Matthew S.

    2005-03-01

    The next generation of electronic devices will be developed at the nanoscale and molecular level, where quantum mechanical effects are observed. These effects must be accounted for in the design process for such small devices. One prototypical nanoscale semiconductor device under investigation is a resonant tunneling diode (RTD). Scientists are hopeful the quantum tunneling effects present in an RTD can be exploited to induce and sustain THz frequency current oscillations. To simulate the electron transport within the RTD, the Wigner-Poisson equations are used. These equations describe the time evolution of the electrons distribution within the device. In this paper, this model and a parameter study using this model will be presented. The parameter study involves calculating the steady-state current output from the RTD as a function of an applied voltage drop across the RTD and also calculating the stability of that solution. To implement the parameter study, the computational model was connected to LOCA (Library of Continuation Algorithms), a part of Sandia National Laboratories parallel solver project, Trilinos. Numerical results will be presented.

  8. [Traffic accidents--the national killer].

    PubMed

    Shemer, Joshua

    2004-02-01

    Traffic accidents are the most prevalent cause of death in developed countries between the ages of 1-33 years. In spite of a low motorization level in Israel, the rate of injury per 100,000 residents in Israel (2.8) was higher than in the US (1.8), NZ (1.7), Canada (1.7), Japan (1.3) and most European countries. The worst injuries were among pedestrians; particularly children aged 1-9 years and elderly (70+ years). In the past decade there have been significant advances in trauma care in Israel. Major strides included the foundation of trauma centers in hospitals, the establishment of the National Council for Trauma and the National Center for Trauma and Emergency Medicine Research at the Gertner Institute that coordinates the national trauma registry. One of the primary aims of the registry was to provide data to support decision-makers in setting national policy for accident prevention. The Israeli Police Department provides data on traffic accident victims to the Israeli Central Bureau of Statistics (CBS) which publishes the national figures. In their article in this edition of the journal, Dr Peleg and Dr. Aharonson-Daniel present a grave concern regarding the fact that details of over 50% of hospitalized traffic accident victims were not reported to the CBS by the police, including data on the severely injured casualties. Traffic accidents are a major cause of loss of life and disability, creating a heavy economic burden on the state and the health care system. Hence, the authors recommend establishing a national database which will combine data from medical and other sources and present the complete comprehensive picture of traffic accident injuries. Such a database will improve the decision-making process, providing more focused data to enhance the preparation and dissemination of appropriate injury prevention policies.

  9. 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.

  10. LESSONS LEARNED FROM A RECENT LASER ACCIDENT

    SciTech Connect

    Woods, Michael; /SLAC

    2011-01-26

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

  11. Lessons learned from early criticality accidents

    SciTech Connect

    Malenfant, R.E.

    1996-06-01

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

  12. The medical investigation of airship accidents.

    PubMed

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

    1988-07-01

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

  13. Child protection. Accident prevention: a community approach.

    PubMed

    Roberts, H

    1991-07-01

    Child accidents are the main cause of death and a considerable cause of morbidity in children, as well as anxiety to adults. Attempts to tackle this major health problem have tended to rely on campaigns of education and exhortation; public health strategies remain underdeveloped. Health visitors are well placed to pursue child safety strategies which build on parents' own knowledge and experience. Helen Roberts describes an initiative based not on the question, why did that accident happen? but the more intriguing question of how is it that most parents manage to keep their children safe most of the time and what can we learn from them?

  14. Accident tolerant fuels for LWRs: A perspective

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

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

  15. Czech Republic 20 years after Chernobyl accident.

    PubMed

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

    2008-01-01

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

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

    SciTech Connect

    Kimura, C.Y.

    1993-01-01

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

  17. [Medical protection during radiation accidents: some results and lessons of the Chernobyl accident].

    PubMed

    Legeza, V I; Grebeniuk, A N; Zatsepin, V V

    2011-01-01

    Actions of medical radiation protection of liquidators of consequences of on Chernobyl atomic power station accident are analysed. It is shown, that during the early period of the accident medical protection of liquidators was provided by administration of radioprotectors, means of prophylaxis: of radioactive iodine incorporation and agent for preventing psychological and emotional stress. When carrying out decontamination and regenerative works, preparations which action is caused by increase of nonspecific resistance of an organism were applied. The lessons taken from the results of the Chernobyl accident, have allowed one to improve the system of medical protection and to introduce in practice new highly effective radioprotective agents.

  18. ATMOSPHERIC MODELING IN SUPPORT OF A ROADWAY ACCIDENT

    SciTech Connect

    Buckley, R.; Hunter, C.

    2010-10-21

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

  19. 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.

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

    SciTech Connect

    Hill, Robin L.; Conrady, Matthew M.

    2010-06-30

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

  1. 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.

  2. 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

  3. Remote control continuous mining machine crushing accident data study

    SciTech Connect

    2006-05-11

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

  4. A high-speed hydroplane accident.

    PubMed

    Flaherty, G N

    1975-03-29

    This report records the investigation into a high-speed hydroplane accident in which the driver died. He was ejected head first into the water at 117 to 126 ft/sec (80 to 85 mph), suffering brain damage and a fractured skull. Suggestions are made to minimize the effects of these inevitable crashes. PMID:1143139

  5. School Bus Accidents: Reducing Incidents and Injuries

    ERIC Educational Resources Information Center

    Mahoney, Daniel

    2009-01-01

    The number of children injured in nonfatal school bus accidents annually is more than double the number previously estimated. In Ohio alone, approximately 20,800 children younger than 18 were occupants of school buses that were involved in crashes in 2003 and 2004 (McGeehan 2007). Among those children, most had minor or no injuries. However, there…

  6. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle or taking place on rail transit-controlled property where one or more of the following occurs: (1)...

  7. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle or taking place on rail transit-controlled property where one or more of the following occurs: (1)...

  8. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle or taking place on rail transit-controlled property where one or more of the following occurs: (1)...

  9. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle or taking place on rail transit-controlled property where one or more of the following occurs: (1)...

  10. 49 CFR 659.33 - Accident notification.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Agency § 659.33 Accident notification. (a) The oversight agency must require the rail transit agency to notify the oversight agency within two (2) hours of any incident involving a rail transit vehicle or taking place on rail transit-controlled property where one or more of the following occurs: (1)...

  11. Tragic Car Accident Involves ESO Employees

    NASA Astrophysics Data System (ADS)

    2000-06-01

    Saturday, May 27, turned into a tragic day for ESO. The team installing TIMMI2 at La Silla, went on an excursion to the Elqui valley, 70 km east of the city of La Serena and suffered a serious car accident, crashing against another car driving from the opposite direction.

  12. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 32 National Defense 4 2011-07-01 2011-07-01 false Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  13. [Case Report - Really a diving accident?].

    PubMed

    Fichtner, Andreas

    2015-10-01

    A 17 y old male SCUBA diver presents himself for hospital admission after a suspected diving accident. All clinical signs are favouring the initial diagnosis: loss of leg motor function, paresthesia, disturbed vision and headache. What are your further diagnostic and therapeutic steps? Can you proof the initial diagnosis? What differential diagnoses are relevant or even mimicked? PMID:26510103

  14. [Splenic rupture--a skateboard accident].

    PubMed

    Kruse, P

    1990-03-01

    A 13-year-old boy presented with persisting abdominal pain after a skateboard accident. Primary clinical and laboratory findings disclosed no signs of intra abdominal bleeding. Ultrasound scanning indicated rupture of the spleen which was confirmed by acute exploratory laparotomy.

  15. ANS severe accident program overview & planning document

    SciTech Connect

    Taleyarkhan, R.P.

    1995-09-01

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

  16. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 32 National Defense 4 2010-07-01 2010-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  17. Accident Prevention: A Workers' Education Manual.

    ERIC Educational Resources Information Center

    International Labour Office, Geneva (Switzerland).

    Devoted to providing industrial workers with a greater knowledge of precautionary measures undertaken and enforced by industries for the protection of workers, this safety education manual contains 14 lessons ranging from "The Problems of Accidents during Work" to "Trade Unions and Workers and Industrial Safety." Fire protection, safety equipment…

  18. [Case Report - Really a diving accident?].

    PubMed

    Fichtner, Andreas

    2015-10-01

    A 17 y old male SCUBA diver presents himself for hospital admission after a suspected diving accident. All clinical signs are favouring the initial diagnosis: loss of leg motor function, paresthesia, disturbed vision and headache. What are your further diagnostic and therapeutic steps? Can you proof the initial diagnosis? What differential diagnoses are relevant or even mimicked?

  19. Severe Accident Test Station Activity Report

    SciTech Connect

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

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

  20. [Current situation of accidents in the world].

    PubMed

    Aguilar-Zinser, José Valente

    2010-01-01

    According to the World Health Organization (WHO), the number of traffic accidents is of concern. About 1.2 million people die every year on the roadways and about 20 to 50 million suffer from non-lethal trauma. Countries with low or medium incomes have higher rates of lethality by traffic accidents (21.5 and 19.5 per 100,000 habitants, respectively) than countries with higher incomes (10.3 per 100,000). It is estimated that the cost of traffic accidents in countries that are members of the Organization for Economic Cooperation and Development (OECD), escalate to rates that are between 2-5% of the gross domestic product (GDP). According to data from the health sector in Mexico, these rates are equivalent to 1.3 of GDR The WHO foresees that traffic accident traumas will rise to be the third cause of mortality in 2030. Because of the high complexity of the transport sector, it is necessary that the Transport and Communication Ministry works in a multidisciplinary and intersectorial fashion to ensure that the land transportation systems operate effectively in accordance with national economic development and the quality of life of the Mexican people.

  1. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 32 National Defense 4 2012-07-01 2011-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  2. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 32 National Defense 4 2014-07-01 2013-07-01 true Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  3. 32 CFR 644.532 - Reporting accidents.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 32 National Defense 4 2013-07-01 2013-07-01 false Reporting accidents. 644.532 Section 644.532 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY (CONTINUED) REAL PROPERTY REAL ESTATE HANDBOOK Disposal Clearance of Explosive Hazards and Other Contamination from Proposed Excess...

  4. RECENT LASER ACCIDENTS AT DEPARTMENT OF ENERGY LABORATORIES

    SciTech Connect

    ODOM, CONNON R.

    2007-02-02

    Recent laser accidents and incidents at research laboratories across the Department of Energy complex are reviewed in this paper. Factors that contributed to the accidents are examined. Conclusions drawn from the accident reports are summarized and compared. Control measures that could have been implemented to prevent the accidents will be summarized and compared. Recommendations for improving laser safety programs are outlined and progress toward achieving them are summarized.

  5. Utilization of accident databases and fuzzy sets to estimate frequency of HazMat transport accidents.

    PubMed

    Qiao, Yuanhua; Keren, Nir; Mannan, M Sam

    2009-08-15

    Risk assessment and management of transportation of hazardous materials (HazMat) require the estimation of accident frequency. This paper presents a methodology to estimate hazardous materials transportation accident frequency by utilizing publicly available databases and expert knowledge. The estimation process addresses route-dependent and route-independent variables. Negative binomial regression is applied to an analysis of the Department of Public Safety (DPS) accident database to derive basic accident frequency as a function of route-dependent variables, while the effects of route-independent variables are modeled by fuzzy logic. The integrated methodology provides the basis for an overall transportation risk analysis, which can be used later to develop a decision support system.

  6. Utilization of accident databases and fuzzy sets to estimate frequency of HazMat transport accidents.

    PubMed

    Qiao, Yuanhua; Keren, Nir; Mannan, M Sam

    2009-08-15

    Risk assessment and management of transportation of hazardous materials (HazMat) require the estimation of accident frequency. This paper presents a methodology to estimate hazardous materials transportation accident frequency by utilizing publicly available databases and expert knowledge. The estimation process addresses route-dependent and route-independent variables. Negative binomial regression is applied to an analysis of the Department of Public Safety (DPS) accident database to derive basic accident frequency as a function of route-dependent variables, while the effects of route-independent variables are modeled by fuzzy logic. The integrated methodology provides the basis for an overall transportation risk analysis, which can be used later to develop a decision support system. PMID:19250750

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

    SciTech Connect

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

    2011-06-21

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE

  8. Explaining the road accident risk: weather effects.

    PubMed

    Bergel-Hayat, Ruth; Debbarh, Mohammed; Antoniou, Constantinos; Yannis, George

    2013-11-01

    This research aims to highlight the link between weather conditions and road accident risk at an aggregate level and on a monthly basis, in order to improve road safety monitoring at a national level. It is based on some case studies carried out in Work Package 7 on "Data analysis and synthesis" of the EU-FP6 project "SafetyNet-Building the European Road Safety Observatory", which illustrate the use of weather variables for analysing changes in the number of road injury accidents. Time series analysis models with explanatory variables that measure the weather quantitatively were used and applied to aggregate datasets of injury accidents for France, the Netherlands and the Athens region, over periods of more than 20 years. The main results reveal significant correlations on a monthly basis between weather variables and the aggregate number of injury accidents, but the magnitude and even the sign of these correlations vary according to the type of road (motorways, rural roads or urban roads). Moreover, in the case of the interurban network in France, it appears that the rainfall effect is mainly direct on motorways--exposure being unchanged, and partly indirect on main roads--as a result of changes in exposure. Additional results obtained on a daily basis for the Athens region indicate that capturing the within-the-month variability of the weather variables and including it in a monthly model highlights the effects of extreme weather. Such findings are consistent with previous results obtained for France using a similar approach, with the exception of the negative correlation between precipitation and the number of injury accidents found for the Athens region, which is further investigated. The outlook for the approach and its added value are discussed in the conclusion.

  9. Concussion in Motor Vehicle Accidents: The Concussion Identification Index

    ClinicalTrials.gov

    2016-08-03

    Motor Vehicle Accidents; TBI (Traumatic Brain Injury); Brain Contusion; Brain Injuries; Cortical Contusion; Concussion Mild; Cerebral Concussion; Brain Concussion; Accidents, Traffic; Traffic Accidents; Traumatic Brain Injury With Brief Loss of Consciousness; Traumatic Brain Injury With no Loss of Consciousness; Traumatic Brain Injury With Loss of Consciousness

  10. Developing techniques for cause-responsibility analysis of occupational accidents.

    PubMed

    Jabbari, Mousa; Ghorbani, Roghayeh

    2016-11-01

    The aim of this study was to specify the causes of occupational accidents, determine social responsibility and the role of groups involved in work-related accidents. This study develops occupational accidents causes tree, occupational accidents responsibility tree, and occupational accidents component-responsibility analysis worksheet; based on these methods, it develops cause-responsibility analysis (CRA) techniques, and for testing them, analyzes 100 fatal/disabling occupational accidents in the construction setting that were randomly selected from all the work-related accidents in Tehran, Iran, over a 5-year period (2010-2014). The main result of this study involves two techniques for CRA: occupational accidents tree analysis (OATA) and occupational accidents components analysis (OACA), used in parallel for determination of responsible groups and responsibilities rate. From the results, we find that the management group of construction projects has 74.65% responsibility of work-related accidents. The developed techniques are purposeful for occupational accidents investigation/analysis, especially for the determination of detailed list of tasks, responsibilities, and their rates. Therefore, it is useful for preventing work-related accidents by focusing on the responsible group's duties.

  11. 10 CFR 76.85 - Assessment of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) CERTIFICATION OF GASEOUS DIFFUSION PLANTS Safety § 76.85 Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences...

  12. 10 CFR 76.85 - Assessment of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) CERTIFICATION OF GASEOUS DIFFUSION PLANTS Safety § 76.85 Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences...

  13. 10 CFR 76.85 - Assessment of accidents.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) CERTIFICATION OF GASEOUS DIFFUSION PLANTS Safety § 76.85 Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences...

  14. 10 CFR 76.85 - Assessment of accidents.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) CERTIFICATION OF GASEOUS DIFFUSION PLANTS Safety § 76.85 Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences...

  15. 10 CFR 76.85 - Assessment of accidents.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) CERTIFICATION OF GASEOUS DIFFUSION PLANTS Safety § 76.85 Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences...

  16. 40 CFR 68.168 - Five-year accident history.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 15 2011-07-01 2011-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident...

  17. 40 CFR 68.168 - Five-year accident history.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident...

  18. 29 CFR 1960.70 - Reporting of serious accidents.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 9 2010-07-01 2010-07-01 false Reporting of serious accidents. 1960.70 Section 1960.70... PROGRAMS AND RELATED MATTERS Recordkeeping and Reporting Requirements § 1960.70 Reporting of serious... and catastrophic accident investigation. The summaries shall address the date/time of accident,...

  19. 41 CFR 101-39.407 - Accident records.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... Regulations System FEDERAL PROPERTY MANAGEMENT REGULATIONS AVIATION, TRANSPORTATION, AND MOTOR VEHICLES 39-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.4-Accidents and Claims § 101-39.407 Accident records. If GSA's records... 41 Public Contracts and Property Management 2 2014-07-01 2012-07-01 true Accident records....

  20. 43 CFR 15.13 - Report of accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 43 Public Lands: Interior 1 2012-10-01 2011-10-01 true Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as...

  1. 43 CFR 15.13 - Report of accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as...

  2. 43 CFR 15.13 - Report of accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 43 Public Lands: Interior 1 2011-10-01 2011-10-01 false Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as...

  3. 43 CFR 15.13 - Report of accidents.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 43 Public Lands: Interior 1 2013-10-01 2013-10-01 false Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as...

  4. 40 CFR 68.168 - Five-year accident history.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 16 2012-07-01 2012-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident...

  5. 40 CFR 68.168 - Five-year accident history.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 16 2013-07-01 2013-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident...

  6. 40 CFR 68.168 - Five-year accident history.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 16 2014-07-01 2014-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident...

  7. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 1 2013-10-01 2013-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident...

  8. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

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

  9. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 1 2011-10-01 2011-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident...

  10. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

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

  11. 32 CFR 636.13 - Traffic accident investigation reports.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ....29 of this subchapter: (a) Military Police at Fort Stewart/Hunter Army Airfield installations will record traffic accident investigations on DA Form 3946 (Military Police Traffic Accident Report) and DA Form 3975 (Military Police Report). (b) All privately owned motor vehicle accidents on Fort Stewart...

  12. 32 CFR 636.13 - Traffic accident investigation reports.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ....29 of this subchapter: (a) Military Police at Fort Stewart/Hunter Army Airfield installations will record traffic accident investigations on DA Form 3946 (Military Police Traffic Accident Report) and DA Form 3975 (Military Police Report). (b) All privately owned motor vehicle accidents on Fort Stewart...

  13. 48 CFR 52.236-13 - Accident Prevention.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Accident Prevention. 52....236-13 Accident Prevention. As prescribed in 36.513, insert the following clause: Accident Prevention... the Secretary of Labor at 29 CFR part 1926 and 29 CFR part 1910; and (3) Ensure that any...

  14. 49 CFR 382.303 - Post-accident testing.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 5 2014-10-01 2014-10-01 false Post-accident testing. 382.303 Section 382.303... ALCOHOL USE AND TESTING Tests Required § 382.303 Post-accident testing. (a) As soon as practicable... functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who...

  15. 49 CFR 840.3 - Notification of railroad accidents.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 7 2011-10-01 2011-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center...

  16. 49 CFR 840.3 - Notification of railroad accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 7 2014-10-01 2014-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center...

  17. 49 CFR 840.3 - Notification of railroad accidents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 7 2012-10-01 2012-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center...

  18. 49 CFR 382.303 - Post-accident testing.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 5 2012-10-01 2012-10-01 false Post-accident testing. 382.303 Section 382.303... ALCOHOL USE AND TESTING Tests Required § 382.303 Post-accident testing. (a) As soon as practicable... functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who...

  19. 49 CFR 840.3 - Notification of railroad accidents.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center...

  20. 49 CFR 840.3 - Notification of railroad accidents.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 7 2013-10-01 2013-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center...

  1. Post-Traumatic Stress Disorders in Reactions to Car Accidents.

    ERIC Educational Resources Information Center

    Pasahow, Robert J.

    This paper explains the most frequent psychological symptom that a car accident victim experiences and describes the nature of an anxiety and avoidant reaction to being in a car following an accident. The description of these responses is based on clinical and in-vivo observations from the treatment of more than 450 cases. Accident victim…

  2. 49 CFR 382.303 - Post-accident testing.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 5 2013-10-01 2013-10-01 false Post-accident testing. 382.303 Section 382.303... ALCOHOL USE AND TESTING Tests Required § 382.303 Post-accident testing. (a) As soon as practicable... functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who...

  3. 46 CFR 4.03-1 - Marine casualty or accident.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident...

  4. [Fatal electric arc accidents due to high voltage].

    PubMed

    Strauch, Hansjürg; Wirth, Ingo

    2004-01-01

    The frequency of electric arc accidents has been successfully reduced owing to preventive measures taken by the professional association. However, the risk of accidents has continued to exist in private setting. Three fatal electric arc accidents caused by high voltage are reported with reference to the autopsy findings.

  5. 33 CFR 173.55 - Report of casualty or accident.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Report of casualty or accident. 173.55 Section 173.55 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) BOATING SAFETY VESSEL NUMBERING AND CASUALTY AND ACCIDENT REPORTING Casualty and Accident Reporting § 173.55 Report of casualty or...

  6. 14 CFR 294.40 - Aircraft accident liability insurance requirements.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 14 Aeronautics and Space 4 2012-01-01 2012-01-01 false Aircraft accident liability insurance....40 Aircraft accident liability insurance requirements. No Canadian charter air taxi operator shall engage in charter air service unless such carrier has and maintains in effect aircraft accident...

  7. 43 CFR 15.13 - Report of accidents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 43 Public Lands: Interior 1 2014-10-01 2014-10-01 false Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as...

  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. 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

  10. 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.

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

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

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

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

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

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

  13. 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

  14. Nuclear Facilities Fire Accident Model

    1999-09-01

    4. NATURE OF PROBLEM SOLVED FIRAC predicts fire-induced flows, thermal and material transport, and radioactive and nonradioactive source terms in a ventilation system. It is designed to predict the radioactive and nonradioactive source terms that lead to gas dynamic, material transport, and heat transfer transients. FIRAC's capabilities are directed toward nuclear fuel cycle facilities and the primary release pathway, the ventilation system. However, it is applicable to other facilities and can be used to modelmore » other airflow pathways within a structure. The basic material transport capability of FIRAC includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, and gas dynamics are also simulated. A ventilation system model includes elements such as filters, dampers, ducts, and blowers connected at nodal points to form networks. A zone-type compartment fire model is incorporated to simulate fire-induced transients within a facility. 5. METHOD OF SOLUTION FIRAC solves one-dimensional, lumped-parameter, compressible flow equations by an implicit numerical scheme. The lumped-parameter method is the basic formulation that describes the gas dynamics system. No spatial distribution of parameters is considered in this approach, but an effect of spatial distribution can be approximated by noding. Network theory, using the lumped parameter method, includes a number of system elements, called branches, joined at certain points, called nodes. Ventilation system components that exhibit flow resistance and inertia, such as dampers, ducts, valves, and filters, and those that exhibit flow potential, such as blowers, are located within the branches of the system. The connection points of branches are nodes for components that have finite volumes, such as rooms, gloveboxes, and plenums, and for boundaries where the volume is practically infinite. All internal nodes, therefore, possess some

  15. Two Simple Formulas for Evaluating the Lower Bound of the Impact Velocity in Vehicle-pedestrian Accidents.

    PubMed

    Zou, Tiefang; Zhang, Yonggang; Yin, Ruoyu

    2016-07-01

    Formulas for evaluating the lower bound of the impact velocity are valuable in vehicle-pedestrian accident reconstruction. The theory of classical mechanics and four hypotheses were employed to derive formulas; the research results and simulation/accident tests were employed to validate their feasibility. Then, two simple formulas were developed according to the distance between the rest positions of the vehicle and the pedestrian and the flight-phase distance. The results showed that the evaluated results by the two proposed formulas are inferior to the existing results. The influence of a roadside step on the impact velocity, which decreased with an increase in the flight-phase distance and a reduction in the road slope, was evaluated. Based on a real accident, the study concludes that the lower bound can be easily obtained with the proposed formulas, which can be used to determine the evaluated impact velocity during simulations. PMID:27364273

  16. Two Simple Formulas for Evaluating the Lower Bound of the Impact Velocity in Vehicle-pedestrian Accidents.

    PubMed

    Zou, Tiefang; Zhang, Yonggang; Yin, Ruoyu

    2016-07-01

    Formulas for evaluating the lower bound of the impact velocity are valuable in vehicle-pedestrian accident reconstruction. The theory of classical mechanics and four hypotheses were employed to derive formulas; the research results and simulation/accident tests were employed to validate their feasibility. Then, two simple formulas were developed according to the distance between the rest positions of the vehicle and the pedestrian and the flight-phase distance. The results showed that the evaluated results by the two proposed formulas are inferior to the existing results. The influence of a roadside step on the impact velocity, which decreased with an increase in the flight-phase distance and a reduction in the road slope, was evaluated. Based on a real accident, the study concludes that the lower bound can be easily obtained with the proposed formulas, which can be used to determine the evaluated impact velocity during simulations.

  17. 49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    .... For information regarding the types of documents routinely issued by the Board, see 49 CFR part 801... public accident investigation dockets, and accident database data. 837.3 Section 837.3 Transportation... investigation dockets, and accident database data. (a) Demands for material contained in the NTSB's...

  18. 49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    .... For information regarding the types of documents routinely issued by the Board, see 49 CFR part 801... public accident investigation dockets, and accident database data. 837.3 Section 837.3 Transportation... investigation dockets, and accident database data. (a) Demands for material contained in the NTSB's...

  19. 49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    .... For information regarding the types of documents routinely issued by the Board, see 49 CFR part 801... public accident investigation dockets, and accident database data. 837.3 Section 837.3 Transportation... investigation dockets, and accident database data. (a) Demands for material contained in the NTSB's...

  20. 49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    .... For information regarding the types of documents routinely issued by the Board, see 49 CFR part 801... public accident investigation dockets, and accident database data. 837.3 Section 837.3 Transportation... investigation dockets, and accident database data. (a) Demands for material contained in the NTSB's...