WorldWideScience

Sample records for accident consequence analysis

  1. Hanford Waste Tank Bump Accident and Consequence Analysis

    Energy Technology Data Exchange (ETDEWEB)

    BRATZEL, D.R.

    2000-06-20

    This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks.

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

    CERN Document Server

    Carro, C A

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

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

  5. Chernobyl accident and its consequences

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-06-01

    The paper concerns the Chernobyl reactor accident, with emphasis on the design of the RBMK reactor and nuclear safety. A description is given of the Chernobyl nuclear power plant, including details of the RMBK reactor and safety systems. Comments on the design of the RBMK by UK experts prior to the accident are summarized, along with post-accident design changes to improve RBMK safety. Events of the Chernobyl accident are described, as well as design deficiencies highlighted by the accident. Differences between the USSR and UK approaches to nuclear safety are commented on. Finally source terms, release periods and environmental consequences are briefly discussed.

  6. Chernobyl accident and its consequences

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.; Bonell, P.G.; Hicks, D.

    1987-01-01

    The USSR power reactor programme is first described. The reasons for the accident at the Chernobyl-4 RBMK nuclear reactor on 26 April 1986, the sequence of events that took place, and the immediate and long-term consequences are considered. A description of the RBMK-type reactors is given and the design changes resulting from the experience of the accident are explained. The source terms describing the details of the radioactivity release associated with the accident and the environmental consequences are covered in the last two sections of the report. Throughout the text comments referring to the UK Nuclear Installations Inspectorate Safety assessment principles have been inserted. (U.K.).

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

    Energy Technology Data Exchange (ETDEWEB)

    Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1998-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [Univ. of New Mexico, Albuquerque, NM (United States); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

    1995-01-01

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

  13. Review of methodology for accident consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Strenge, D.L.; Soldat, J.K.; Watson, E.C.

    1978-09-01

    This report reviews current methodologies for reactor accident consequence analysis and describes areas where modifications are warranted. Methodologies reviewed are: (1) Models in Regulatory Guides 1.109, 1.111 and 1.113 used for evaluation of compliance with 10 CFR 50 Appendix I; (2) Models in Regulatory Guides used for evaluation of consequences from accidents of Classes 3-8; (3) Models for evaluation of Class 9 accidents presented in the Reactor Safety Study; and (4) Models in the Liquid Pathway Generic Study. The review is designed to aid in the ultimate goal of selection of a comprehensive set of models to extend the Class 9 methodology of the Reactor Safety Study to the analysis of Classes 3-8 accidents.

  14. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  15. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  16. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

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

    2001-05-01

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

  19. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

  20. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States)] [and others

    1995-01-01

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

  1. Medical consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Galstyan I.A.

    2015-12-01

    Full Text Available Aim: to study the long-term effects of acute radiation syndrome (ARS, developed at the victims of the Chernobyl accident. Material and Methods. 237 people were exposed during the accident, 134 of them were diagnosed with ARS. Dynamic observation implies a thorough annual examination in a hospital. Results. In the first 1.5-2 years after the ARS mean group indices of peripheral blood have returned to normal. However, many patients had transient expressed moderate cytopenias. Granulocytopenia, thrombocytopenia, lymphopenia and erythropenia were the most frequently observed things during the first 5 years after the accident. After 5 years their occurences lowered. In 11 patients the radiation cataract was detected. A threshold dose for its development is a dose of 3.2 Gy Long-term effects of local radiation lesions (LRL range from mild skin figure smoothing to a distinct fibrous scarring, contractures, persistently recurrent late radiation ulcers. During all years of observation we found 8 solid tumors, including 2 thyroid cancers. 5 hematologic diseases were found. During 29 years 26 ARS survivors died of various causes. Conclusion. The health of ones with long-term ARS effects is determined by the evolution of the LRL effects on skin, radiation cataracts, hema-tological diseases and the accession of of various somatic diseases, not caused by radiation.

  2. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2001-07-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  3. Health effects models for nuclear power plant accident consequence analysis. Part 1, Introduction, integration, and summary: Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S. [Harvard School of Public Health, Boston, MA (United States); Abrahmson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Inhalation Toxicology Research Inst., Albuquerque, NM (United States); Gilbert, E.S. [Battelle Pacific Northwest Lab., Richland, WA (United States)

    1993-10-01

    This report is a revision of NUREG/CR-4214, Rev. 1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. This revision has been made to incorporate changes to the Health Effects Models recommended in two addenda to the NUREG/CR-4214, Rev. 1, Part 11, 1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR, ICRP and NAS/NRC (BEIR V). The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes are considered. Linear and linear-quadratic models are recommended for estimating the risks of seven types of cancer in adults - leukemia, bone, lung, breast, gastrointestinal, thyroid, and ``other``. For most cancers, both incidence and mortality are addressed. Five classes of genetic diseases -- dominant, x-linked, aneuploidy, unbalanced translocations, and multifactorial diseases are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk.

  4. First international workshop on severe accidents and their consequences. [Chernobyl Accident

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

    An international workshop on past severe nuclear accidents and their consequences was held in Dagomys region of Sochi, USSR on October 30--November 3, 1989. The plan of this meeting was approved by the USSR Academy of Sciences and by the USSR State Committee of the Utilization of Atomic Energy. The meeting was held under the umbrella of the ANS-SNS agreement of cooperation. Topics covered include analysis of the Chernobyl accident, safety measures for RBMK type reactors and consequences of the Chernobyl accident including analysis of the ecological, genetic and psycho-social factors. Separate reports are processed separately for the data bases. (CBS)

  5. The Chernobyl accident consequences; Consequences de l'accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  6. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S. (Harvard Univ., Boston, MA (USA). School of Public Health)

    1990-01-01

    This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, models are included for assessing the risks of several nonlethal early and continuing effects -- including prodromal vomiting and diarrhea, hypothyroidism and radiation thyroiditis, skin burns, reproductive effects, and pregnancy losses. Linear and linear-quadratic models are recommended for estimating cancer risks. Parameters are given for analyzing the risks of seven types of cancer in adults -- leukemia, bone, lung, breast, gastrointestinal, thyroid, and other.'' The category, other'' cancers, is intended to reflect the combined risks of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus and cervix. Models of childhood cancers due to in utero exposure are also developed. For most cancers, both incidence and mortality are addressed. The models of cancer risk are derived largely from information summarized in BEIR III -- with some adjustment to reflect more recent studies. 64 refs., 18 figs., 46 tabs.

  7. Consequences of severe nuclear accidents in Europe

    Science.gov (United States)

    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

  8. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

  9. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    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.

  10. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Jow, H.N. (Sandia National Labs., Albuquerque, NM (USA))

    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.

  11. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    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. Nuclear Accidents: Consequences for Human, Society and Energy Sector

    Directory of Open Access Journals (Sweden)

    L. A. Bolshov

    2016-01-01

    Full Text Available The article examines radiation and hygienic regulations with regard to the elimination of consequences of the Chernobyl NPP accident in the context of relationships with other aspects, primarily socio-economic and political factors. This experience is reasonable to take into account when defining criteria in other regulatory fields, for example, in radioactive waste classification and remediation of areas. The article presents an analysis of joint features and peculiarities of nuclear accidents in the industry and energy sectors. It is noted that the scale of global consequences of the Chernobyl NPP accident is defined by the large-scale release of radioactivity into the environment, as well as an affiliation of the nuclear installation with the energy sector. Large-scale radiation accidents affect the most diverse spheres of human activities, what, in its turn, evokes the reverse reaction from the society and its institutions, including involvement of political means of settlement. If the latter is seeing for criteria that are scientifically justified and feasible, then the preconditions for minimizing socio-economic impacts are created. In other cases, political decisions, such as nuclear units’ shutdown and phasing out of nuclear energy, appear to be an economic price which society, as a whole and a single industry sector, pay to compensate the negative public response. The article describes fundamental changes in approaches to ensure nuclear and radiation safety that occurred after the Chernobyl NPP accident. Multiple and negative consequences of the Chernobyl accident for human and society are balanced to some extent by a higher level of operational safety, emergency preparedness, and life-cycle safety. The article indicates that harmonization and ensuring consistency of regulations that involve different aspects of nuclear and radiation safety are important to implement practical solutions to the nuclear legacy problems. The

  13. 大型 LNG 储罐泄漏事故后果分析%Consequence Analysis of Leakage Accident of Large LNG Storage Tank

    Institute of Scientific and Technical Information of China (English)

    俞志东; 吴建林

    2016-01-01

    Natural gas is a kind of high quality ,high efficiency ,clean energy ,Which is widely used in various industries .In order to use natural gas Conveniently ,some coastal cities in our country have built many large LNG storage tanks .Liquefied natural gas is the flammable and explosive dangerous goods ,so once a tank leaks ,it may cause very serious consequences .The PHAST software ,once being input the real scene at the time of an accident ,can calculate and analyze the diffusion ,the flash fire ,the fire disaster ,or the explosion immediately , quickly determining the leakage ,diffusion ,combustion and explosion hazard area ,and measuring the personnel death radius ,serious injury radius and minor injury radius caused by the fire thermal radiation and explosion pressure .Analysis of the accident can help receiving sta-tions intuitively understand the degree of danger after the accident so as to take corresponding measures .%天然气是一种优质、高效的清洁能源,广泛地应用于各个行业中。我国部分沿海城市已建造了大型 LNG 接收站,运用PHAST 软件,输入发生事故时的真实场景,可对 LNG 储罐泄漏事故后果进行计算和分析,从而确定泄漏、扩散、燃烧和爆炸等危险区域范围。分析结果能帮助接收站直观了解发生事故后的危险程度,进而能够针对性地采取相应的安全措施。

  14. [The Fukushima nuclear accident: consequences for Japan and for us].

    Science.gov (United States)

    Grosche, B

    2013-04-01

    The Fukushima accident was the consequence of a preceding 2-fold natural catastrophe: the earth quake of 11 March 2011 and the subsequent tsunami. Due to favourable winds and to evacuation measures the radiation exposure to the general population in Japan as a whole and with some exceptions in the region outside the evacuation zone, too, was low. In this article the attempt is made to give an estimate of health consequences to the public. This is based upon WHO's dose estimates, knowledge of the consequences of the Chernobyl accident, of the atmospheric nuclear bomb testing in Kazakhstan and on the risk of childhood leukaemia after low dose radiation exposure. For Germany, there was no radiation threat due to the accident. Nonetheless, the events in Japan made clear that the rules and standards that were developed for the case of a reactor accident need to be revised.

  15. Accident Tolerant Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    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

  16. Accident tolerant fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Laboratory; Chichester, Heather [Idaho National Laboratory; Johns, Jesse [Texas A& M University; Teague, Melissa [Idaho National Laboratory; Tonks, Michael Idaho National Laboratory; Youngblood, Robert [Idaho National Laboratory

    2014-09-01

    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

  17. The Quota Simulation Analysis in Accident Consequence of Ethanol Tank Area%乙醇贮罐区事故后果定量模拟分析

    Institute of Scientific and Technical Information of China (English)

    苗金明; 刘倩倩; 张宪金

    2012-01-01

    This paper systematically analyzes the major risk of the ethanol tank area, and summarizes the possible consequences of the accident: tank poor fire, vapor cloud explosion (VCE), boiling liquid expanding vapor explosion (BLEVE). Also mathematical models of quan- titative simulation analysis and an applied methodology for simulating the consequences is presen- ted, which is conductive for overall layout of alcohol tank, security design, risk analysis, and accident prevention measures.%通过对乙醇罐区的危险特性进行分析,指出乙醇的事故后果主要为池火灾、蒸气云爆炸、沸腾液体扩展为蒸气爆炸;给出了乙醇事故后果的定量模拟分析的数学模型,同时进行了实例模拟,对指导乙醇贮罐区的总体布局、贮罐区安全设计、贮罐区危险性分析及事故预防措施均有一定参考价值。

  18. Uncertainties in offsite consequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Young, M.L.; Harper, F.T.; Lui, C.H.

    1996-03-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequences from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the U.S. Nuclear Regulatory Commission and the European Commission began co-sponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables using a formal expert judgment elicitation and evaluation process. This paper focuses on the methods used in and results of this on-going joint effort.

  19. ASSESSMENT OF THE FUKUSIMA NUCLEAR POWER PLANT ACCIDENT CONSEQUENCES BY THE POPULATION IN THE FAR EAST

    Directory of Open Access Journals (Sweden)

    G. V. Arkhangelskaya

    2012-01-01

    Full Text Available The article analyzes the attitude of the population in the five regions of the Far East to the consequences of the accident at the Fukushimai nuclear power plant, as well as the issues of informing about the accident. The analysis of public opinion is based on the data obtained by anonymous questionnaire survey performed in November 2011. In spite of the rather active informing and objective information on the absence of the contamination, most of the population of the Russian Far East believes that radioactive contamination is presented in the areas of their residence, and the main cause of this contamination is the nuclear accident in Japan.

  20. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

  1. Accident investigation and analysis

    NARCIS (Netherlands)

    Kampen, J. van; Drupsteen, L.

    2013-01-01

    Many organisations and companies take extensive proactive measures to identify, evaluate and reduce occupational risks. However, despite these efforts things still go wrong and unintended events occur. After a major incident or accident, conducting an accident investigation is generally the next ste

  2. Atmospheric transport patterns and possible consequences for the European North after a nuclear accident.

    Science.gov (United States)

    Baklanov, A; Mahura, A; Jaffe, D; Thaning, L; Bergman, R; Andres, R

    2002-01-01

    The main purpose of this study is to examine possible impacts and consequences of a hypothetical accident at the Kola nuclear plant in north-west Russia on different geographical regions: Scandinavia, central Europe, European FSU and Taymyr. The period studied is 1991-1996. An isentropic trajectory model has been used to calculate forward trajectories that originated over the nuclear accident region. Atmospheric transport patterns were identified using the isentropic trajectories and a cluster analysis technique. From the trajectory model results, a number of cases were chosen for examination in detail using more complete transport models. For this purpose, the models MATHEW/ADPIC, DERMA and a newly developed FOA Random Displacement Model have been used to simulate the radionuclide transport and contamination in the case of a nuclear accident and their results have been compared with those of the trajectory modelling. Estimation of the long-term consequences for populations after an accident has been performed for several specific dates by empirical models and correlation between fallout and doses to humans on the basis of the Chernobyl accident exposures in Scandinavia.

  3. Atmospheric transport patterns and possible consequences for the European North after a nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Baklanov, A. E-mail: alb@dmi.min.dk; Mahura, A.; Jaffe, D.; Thaning, L.; Bergman, R.; Andres, R

    2002-07-01

    The main purpose of this study is to examine possible impacts and consequences of a hypothetical accident at the Kola nuclear plant in north-west Russia on different geographical regions: Scandinavia, central Europe, European FSU and Taymyr. The period studied is 1991-1996. An isentropic trajectory model has been used to calculate forward trajectories that originated over the nuclear accident region. Atmospheric transport patterns were identified using the isentropic trajectories and a cluster analysis technique. From the trajectory model results, a number of cases were chosen for examination in detail using more complete transport models. For this purpose, the models MATHEW/ADPIC, DERMA and a newly developed FOA Random Displacement Model have been used to simulate the radionuclide transport and contamination in the case of a nuclear accident and their results have been compared with those of the trajectory modelling. Estimation of the long-term consequences for populations after an accident has been performed for several specific dates by empirical models and correlation between fallout and doses to humans on the basis of the Chernobyl accident exposures in Scandinavia.

  4. Radioactive materials transport accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    McSweeney, T.I.; Maheras, S.J.; Ross, S.B. [Battelle Memorial Inst. (United States)

    2004-07-01

    Over the last 25 years, one of the major issues raised regarding radioactive material transportation has been the risk of severe accidents. While numerous studies have shown that traffic fatalities dominate the risk, modeling the risk of severe accidents has remained one of the most difficult analysis problems. This paper will show how models that were developed for nuclear spent fuel transport accident analysis can be adopted to obtain estimates of release fractions for other types of radioactive material such as vitrified highlevel radioactive waste. The paper will also show how some experimental results from fire experiments involving low level waste packaging can be used in modeling transport accident analysis with this waste form. The results of the analysis enable an analyst to clearly show the differences in the release fractions as a function of accident severity. The paper will also show that by placing the data in a database such as ACCESS trademark, it is possible to obtain risk measures for transporting the waste forms along proposed routes from the generator site to potential final disposal sites.

  5. Final report of the accident phenomenology and consequence (APAC) methodology evaluation. Spills Working Group

    Energy Technology Data Exchange (ETDEWEB)

    Brereton, S.; Shinn, J. [Lawrence Livermore National Lab., CA (United States); Hesse, D [Battelle Columbus Labs., OH (United States); Kaninich, D. [Westinghouse Savannah River Co., Aiken, SC (United States); Lazaro, M. [Argonne National Lab., IL (United States); Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The Spills Working Group was one of six working groups established under the Accident Phenomenology and Consequence (APAC) methodology evaluation program. The objectives of APAC were to assess methodologies available in the accident phenomenology and consequence analysis area and to evaluate their adequacy for use in preparing DOE facility safety basis documentation, such as Basis for Interim Operation (BIO), Justification for Continued Operation (JCO), Hazard Analysis Documents, and Safety Analysis Reports (SARs). Additional objectives of APAC were to identify development needs and to define standard practices to be followed in the analyses supporting facility safety basis documentation. The Spills Working Group focused on methodologies for estimating four types of spill source terms: liquid chemical spills and evaporation, pressurized liquid/gas releases, solid spills and resuspension/sublimation, and resuspension of particulate matter from liquid spills.

  6. Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA))

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.

  7. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. RADIATION-HYGIENIC AND MEDICAL CONSEQUENCES OF THE СHERNOBYL ACCIDENT: RESULTS AND PROGNOSIS

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2011-01-01

    Full Text Available An article is devoted to the analysis of the radiation situation in the dynamics during the years since the accident at the Chernobyl nuclear power plant in 1986. Data on the scope of activities fulfilled for the assessment of the territories radioactive contamination levels and foodstuffs contamination levels, on the values of the exposure doses for the population living on the contaminated territories, on the medical and socio-psychological consequences of the Chernobyl accident is presented. Basic norms and principles, used during the protective measures development and introduction, are considered, their effectiveness is demonstrated. Mistakes emerged during protective measures implementation are analyzed, the prognosis of the population exposure dose values for the 70-year period since the accident and main directions of activities for the contaminated territories remediation and normal life conditions restoration for the population at these territories are presented.

  9. Health effects models for nuclear power plant accident consequence analysis. Modification of models resulting from addition of effects of exposure to alpha-emitting radionuclides: Revision 1, Part 2, Scientific bases for health effects models, Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Abrahamson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (United States). Inhalation Toxicology Research Inst.; Gilbert, E.S. [Pacific Northwest Lab., Richland, WA (United States)

    1993-05-01

    The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled {open_quotes}Health Effects Models for Nuclear Power Plant Consequence Analysis{close_quotes}, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled {open_quotes}Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models,{close_quotes} was published in 1991 as NUREG/CR-4214, Rev. 1, Part 2, Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure to low-LET beta and gamma radiations is a reasonable extension of the health effects model.

  10. Practical approaches in accident analysis

    Science.gov (United States)

    Stock, M.

    An accident analysis technique based on successive application of structural response, explosion dynamics, gas cloud formation, and plant operation failure mode models is proposed. The method takes into account the nonideal explosion characteristic of a deflagration in the unconfined cloud. The resulting pressure wave differs significantly from a shock wave and the response of structures like lamp posts and walls can differ correspondingly. This gives a more realistic insight into explosion courses than a simple TNT-equivalent approach.

  11. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-07-15

    In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO{sub 2} matrix

  12. Severe accident analysis using dynamic accident progression event trees

    Science.gov (United States)

    Hakobyan, Aram P.

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

  13. Prevention of "simple accidents at work" with major consequences

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2016-01-01

    broadly. This review identifies gaps in the prevention of simple accidents, relating to safety barriers for risk control and the management processes that need to be in place to deliver those risk controls in a continuingly effective state. The article introduces the ‘‘INFO cards’’ as a tool...... for the systematic observation of hazard sources in order to ascertain whether safety barriers and management deliveries are present. Safety management and safety culture, together with the INFO cards are important factors in the prevention process. The conclusion is that we must look at safety as a part of being...... of prevention or safety methodologies and procedures established for major accidents are applicable to simple accidents. The article goes back to basics about accidents causes, to review the nature of successful prevention techniques and to analyze what have been constraints to getting this knowledge used more...

  14. Radioecological and dosimetric consequences of Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l`accident de Tchernobyl en France

    Energy Technology Data Exchange (ETDEWEB)

    Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph

    1997-12-31

    After ten years and the taking in account of numerous data, it can be affirmed that the dosimetric consequences of Chernobyl accident will have been limited in France. for the period 1986-2046, the individual middle efficient dose commitment, for the area the most reached by depositing is inferior to 1500 {mu}Sv, that represents about 1% of middle natural exposure in the same time. but mountains and forests can have more important surface activities than in plain. Everywhere else, it can be considered that the effects of Chernobyl accident are disappearing. the levels of cesium 137 are now often inferior to what they were before the accident. (N.C.)

  15. 汽油储罐化学爆炸事故后果模拟分析%Simulation Analysis of the Consequence of Chemical Gasoline Tank Explosion Accident

    Institute of Scientific and Technical Information of China (English)

    张啸

    2014-01-01

    一直以来,汽油储罐化学爆炸事故模型被用来评审加油站的安全评价报告,所以本文通过实际模拟计算来分析汽油储罐化学爆炸事故模型,以方便加油站进行安全管理。%Chemical gasoline tank explosion accident model has been used to review the safety assessment report of the gas station for a long time. So this article analyzes the chemical gasoline tank explosion model through the actual simulation to facilitate the gas station for safety management.

  16. Assessment of risk, damage and severity of consequences of accident into storage for LPG

    Science.gov (United States)

    Tzenova, Zlatina

    2016-12-01

    In this work an accident scenario in store for LPG is considered and consequences - forming a toxic cloud of vapor, fire and blast are modeled through models built into the software product ALOHA. The risk assessment of contamination with certain concentration is done, provided that it is an accident. Definitions for model mixture and risk assessment using geometric probability are introduced.

  17. Consequences and countermeasures in a nuclear power accident: Chernobyl experience.

    Science.gov (United States)

    Kirichenko, Vladimir A; Kirichenko, Alexander V; Werts, Day E

    2012-09-01

    Despite the tragic accidents in Fukushima and Chernobyl, the nuclear power industry will continue to contribute to the production of electric energy worldwide until there are efficient and sustainable alternative sources of energy. The Chernobyl nuclear accident, which occurred 26 years ago in the former Soviet Union, released an immense amount of radioactivity over vast territories of Belarus, Ukraine, and the Russian Federation, extending into northern Europe, and became the most severe accident in the history of the nuclear industry. This disaster was a result of numerous factors including inadequate nuclear power plant design, human errors, and violation of safety measures. The lessons learned from nuclear accidents will continue to strengthen the safety design of new reactor installations, but with more than 400 active nuclear power stations worldwide and 104 reactors in the Unites States, it is essential to reassess fundamental issues related to the Chernobyl experience as it continues to evolve. This article summarizes early and late events of the incident, the impact on thyroid health, and attempts to reduce agricultural radioactive contamination.

  18. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses; Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Hora, S.C. [Hawaii Univ., Hilo, HI (United States)

    1993-12-01

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community.

  19. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation: Part 2, Scientific bases for health effects models

    Energy Technology Data Exchange (ETDEWEB)

    Abrahamson, S.; Bender, M.; Book, S.; Buncher, C.; Denniston, C.; Gilbert, E.; Hahn, F.; Hertzberg, V.; Maxon, H.; Scott, B.

    1989-05-01

    This report provides dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Two-parameter Weibull hazard functions are recommended for estimating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary and gastrointestinal syndromes -- are considered. Linear and linear-quadratic models are recommended for estimating cancer risks. Parameters are given for analyzing the risks of seven types of cancer in adults -- leukemia, bone, lung, breast, gastrointestinal, thyroid and ''other''. The category, ''other'' cancers, is intended to reflect the combined risks of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus and cervix. Models of childhood cancers due to in utero exposure are also provided. For most cancers, both incidence and mortality are addressed. Linear and linear-quadratic models are also recommended for assessing genetic risks. Five classes of genetic disease -- dominant, x-linked, aneuploidy, unbalanced translocation and multifactorial diseases --are considered. In addition, the impact of radiation-induced genetic damage on the incidence of peri-implantation embryo losses is discussed. The uncertainty in modeling radiological health risks is addressed by providing central, upper, and lower estimates of all model parameters. Data are provided which should enable analysts to consider the timing and severity of each type of health risk. 22 refs., 14 figs., 51 tabs.

  20. Radioecological and dosimetric consequences of the Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l'accident de Tchernobyl en France

    Energy Technology Data Exchange (ETDEWEB)

    Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph. [Inst. de Protection et de Surete Nucleaire, CEA Centre d' Etudes de Fontenay-aux-Roses, 92 (France)

    1997-11-01

    This study has as objective a survey of the radioecological and dosimetric consequences of the Chernobyl accident in France, as well as a prognosis for the years to come. It was requested by the Direction of Nuclear Installation Safety (DSIN) in relation to different organisms which effected measurements after this accident. It is based on the use of combined results of measurements and modelling by means of the code ASTRAL developed at IPSN. Various measurements obtained from five authorities and institutions, were made available, such as: activity of air and water, soil, processed food, agricultural and natural products. However, to achieve the survey still a modelling is needed. ASTRAL is a code for evaluating the ecological consequences of an accident. It allows establishing the correspondence between the soil Remnant Surface Activities (RSA, in Bq.m{sup -2}), the activity concentration of the agricultural production and the individual and collective doses resulting from external and internal exposures (due to inhalation and ingestion of contaminated nurture). The results of principal synthesis documents on the Chernobyl accident and its consequences were also used. The report is structured in nine sections, as follows: 1.Introduction; 2.Objective and methodology; 3.Characterization of radioactive depositions; 4;Remnant surface activities; 5.Contamination of agricultural products and foods; 6.Contamination of natural, semi-natural products and of drinking water; 7.Dosimetric evaluations; 8.Proposals for the environmental surveillance; 9.Conclusion. Finally, after ten years, one concludes that at presentthe dosimetric consequences of the Chernobyl accident in France were rather limited. For the period 1986-2046 the average individual effective dose estimated for the most struck zone is lower than 1500 {mu}Sv, which represents almost 1% of the average natural exposure for the same period. At present, the cesium 137 levels are at often inferior to those recorded

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

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  2. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment.

    Energy Technology Data Exchange (ETDEWEB)

    Thoerring, H.; Liland, A.

    2010-12-15

    This report deals with the environmental consequences in Norway after a hypothetical accident at Sellafield. The investigation is limited to the terrestrial environment, and focus on animals grazing natural pastures, plus wild berries and fungi. Only 137Cs is considered. The predicted consequences are severe, in particular for mutton and goat milk production. (Author)

  3. Consequences of the Chernobyl accident for reindeer husbandry in Sweden

    Directory of Open Access Journals (Sweden)

    Gustaf Åhman

    1990-09-01

    Full Text Available Large parts of the reindeer hearding area in Sweden were contaminated with radioactive caesium from the Chernobyl fallout. During the first year after the accident no food with activity concentrations exceeding 300 Bq/kg was allowed to be sold in Sweden. This meant that about 75% of all reindeer meat produced in Sweden during the autumn and winter 1986/87 were rejected because of too high caesium activités. In May 1987 the maximum level for Cs-137 in reindeer, game and fresh-water fish was raised to 1500 Bq/kg. During the last two year, 1987/88 and 1988/89, about 25% of the slaughtered reindeer has had activities exceeding this limit. The effective long-time halflife or radiocaesium in reindeer after the nuclear weapon tests in the sixties was about 7 years. If this halflife is correct also for the Chernobyl fallout it will take about 35 years before most of the reinder in Sweden are below the current limit 1500 Bq/kg in the winter. However, by feeding the animals uncontaminated food for about two months, many reindeer can be saved for human consumption.

  4. Analysis of Fukushima Daiichi Accident Using HFACS

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Saeed Almheiri [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO{sup 1} and NISA{sup 2} that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident.

  5. RADIOLOGICAL AND MEDICAL CONSEQUENCES OF THE CHERNOBYL ACCIDENT

    Directory of Open Access Journals (Sweden)

    V. G. Bebeshko

    2012-01-01

    Full Text Available From the position of a 25-years’ experience to overcome the health effects of Chernobyl the dynamics of the radiation environment, the first summarizing at the international level (1988, the results of completed research and practical monitoring are analyzed. Cohort of acute radiation syndrome (ARS survivors under medical observation at the S.I. "Research Center for Radiation Medicine of the National Academy of Medical Sciences of Ukraine" is the largest. Within the 25 years the functional state of the major organs and body systems, and metabolic homeostasis for this category of persons were studied, a comprehensive assessment of their health, mental and physical performance were given, and risk factors and peculiarities of stochastic and non-stochastic pathology courses were identified, as well as a system of rehabilitation patients after ARS was developed. ARS survivors are suffering from chronic diseases of internal organs and systems (from 5-7 to 10-12 diagnoses at the same time. A correlation between acute radiation effects and specific HLA phenotypes were revealed. The dynamics of the immune system recovery after irradiation was studied. The role and prognostic value of telomere length and programmed cell death of lymphocytes in the formation of the cellular effects of ionizing radiation were determined for the first time. Differences between spontaneous and radiation-induced acute myeloid leukemias were found. Dose-dependent neuropsychiatric, neurophysiological, neuropsychological and neuroimaging deviations were identified after irradiation at doses above 0.3 Sv. It was shown that the lymphocytes of Chernobyl clean-up workers with doses 350 – 690 mGy can induce "the bystander effect" in the non-irradiated cells even after 19 years after exposure. The rates of cancer incidence and mortality of victims, the lessons and key problems to be solved in the third decade after the Chernobyl accident are considered.

  6. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor (U)

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Olson, R.L.; Hamby, D.M. (Savannah River Lab., Aiken, SC (United States))

    1992-03-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/PBq(2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This paper suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium releases to the atmosphere than from comparable tritium source terms to water pathways.

  7. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs.

  8. An analysis of aircraft accidents involving fires

    Science.gov (United States)

    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.

  9. Safety analysis of surface haulage accidents

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Summary of the consequences for the environment of the Chernobyl accident; Synthese sur les consequences environnementales de l`accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    Ciffroy, P.

    1996-08-01

    The main conclusions on the environmental consequences of the Chernobyl accident in the former Soviet Union can be summarised as follows: the long term radioactive contamination of the environment can essentially be put down to Cs and Sr and, to a lesser degree, transuranic elements. In the short term, the radioactive iodine fall-out plays a fundamental role; in the countries of the former Soviet Union, it is estimated that 29,300 and 10,200 km{sup 2} of the surface area of the land are respectively contaminated by over 185 and 555 kBq.m{sup -2}. Approximately 1,064,000 people live in areas contaminated by more than 185 kBq.m{sup -2}; acute radioactive fall-out effects have occurred in the 30 km exclusion zone, essentially witnessed by the death of numerous conifers. On average, it will take about twenty years for half the Cs to disappear from the top 10 cm of soil; the level of contamination of food products varies greatly according to soil type. However, we can consider that milk, berries and mushrooms were the most critical foods in the years immediately following the accident and that some of the agricultural counter-measures taken have proved very useful in containing the contamination of food products. Because of the massive iodine leakage, the worst affected organ in the body during the months following the accident was the thyroid gland. In the months following the accident, the presence of radioactive elements on the surface of vegetables which were subsequently eaten proved to be the main source of human contamination; after a rapid fall off in external dose received by the population during the first year, it is now decreasing much more slowly. This phenomenon is mainly due to the very long-life of the radioactive caesium in the soil; approximately 90 % of the total internal dose for the 70 years following the accident have already been received by the local population. The external dose level will be reduced fairly slowly and we can assess that

  11. Radioecological consequences of a potential accident during transport of spent nuclear fuel along an Arctic coastline.

    Science.gov (United States)

    Iosjpe, M; Reistad, O; Amundsen, I B

    2009-02-01

    This article presents results pertaining to a risk assessment of the potential consequences of a hypothetical accident occurring during the transportation by ship of spent nuclear fuel (SNF) along an Arctic coastline. The findings are based on modelling of potential releases of radionuclides, radionuclide transport and uptake in the marine environment. Modelling work has been done using a revised box model developed at the Norwegian Radiation Protection Authority. Evaluation of the radioecological consequences of a potential accident in the southern part of the Norwegian Current has been made on the basis of calculated collective dose to man, individual doses for the critical group, concentrations of radionuclides in seafood and doses to marine organisms. The results of the calculations indicate a large variability in the investigated parameters above mentioned. On the basis of the calculated parameters the maximum total activity ("accepted accident activity") in the ship, when the parameters that describe the consequences after the examined potential accident are still in agreement with the recommendations and criterions for protection of the human population and the environment, has been evaluated.

  12. Consequences to health of the Chernobyl accident; Helbredsmaessige konsekvenser af reaktorulykken i Tjernobyl

    Energy Technology Data Exchange (ETDEWEB)

    Sewerin, I. [Royal Dental College, Dept. of Radiology, Copenhagen (Denmark)

    2001-07-01

    The Chernobyl accident in 1986 has been and still is the subject of great interest. Journalistic reports often contain exaggerations and undocumented statements and much uncertainty about the true consequences of the accident prevails in the population. This article reviews the current literature with the focus on reports from official commissions and documentation in the form of controlled studies. The fatal deterministic consequences comprise about 30 victims. The most important outcome is a marked increase in the incidence of thyroid cancer in children and adolescents in the most heavily contaminated area. Furthermore, pronounced psychosocial problems are dominant in the population of the contaminated area. Other significant and documented health consequences are not seen. (au)

  13. HTGR severe accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Cold Vacuum Drying facility design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    2000-08-08

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

  15. Two decades of radiological accidents direct causes, roots causes and consequences

    Directory of Open Access Journals (Sweden)

    Rozental Jose de Julio

    2002-01-01

    Full Text Available Practically all Countries utilize radioisotopes in medicine, industry, agriculture and research. The extent to which ionizing radiation practices are employed varies considerably, depending largely upon social and economic conditions and the level of technical skills available in the country. An overview of the majority of practices and the associated hazards will be found in the Table IV to VII of this document. The practices in normal and abnormal operating conditions should follow the basic principles of radiation protection and the Safety of Radiation Sources, considering the IAEA Radiation Protection and the Safety of Radiation Sources, Safety Series 120 and the IAEA Recommendation of the Basic Safety Standards for Radiation Protection, Safety Series Nº 115. The Standards themselves underline the necessity to be able to predict the radiological consequences of emergency conditions and the investigations that should need to be done. This paper describes the major accidents that had happened in the last two decades, provides a methodology for analyses and gives a collection of lessons learned. This will help the Regulatory Authority to review the reasons of vulnerabilities, and to start a Radiation safety and Security Programme to introduce measurescapable to avoid the recurrence of similar events. Although a number of accidents with fatalities have caught the attention of the public in recent year, a safety record has accompanied the widespread use of radiation sources. However, the fact that accidents are uncommon should not give grounds for complacency. No radiological accident is acceptable. From a radiation safety and security of the sources standpoint, accident investigation is necessary to determine what happened, why, when, where and how it occurred and who was (were involved and responsible. The investigation conclusion is an important process toward alertness and feedback to avoid careless attitudes by improving the comprehension

  16. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  17. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    Energy Technology Data Exchange (ETDEWEB)

    O`Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-12-31

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 {times} 10{sup {minus}3} person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 {times} 10{sup {minus}2} person-Sv/Pbq (2.4 {times} 10{sup {minus}4} person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river.

  18. The accident simulation and consequence analysis of the hydrogen refueling station leakage%加氢站氢气泄漏事故模拟及后果分析

    Institute of Scientific and Technical Information of China (English)

    杨灿剑; 付晋

    2011-01-01

    Aim at the safety of hydrogen refueling station, the theoretical model analysis and numerical simulation was carried out to simulate the accident and consequence. The concentration contours were obtained by the in house MATLAB gauss diffusion code. The influence on hydrogen diffusion of wind speed in the surroundings was analyzed. That is, the hazardous area decreased when the wind speed up. The CFD code Fluent was adopted to build a 2D full-scale hydrogen leakage in the hydrogen refueling station. The results showed that, under calm condition, the horizontal and vertical hydrogen diffusions are very fast and tend to accumulate forming explosive gas cloud. But under 10 m/s wind, the escaping gas is driven, blown away and diluted and it is hard to gather. The blast area is limited next to the leaking source. The wind is not benefit to the hydrogen diffusion but is good for safety.%针对加氢站安全,通过理论模型分析和数值模拟两种方法,对其开展事故模拟和后果分析.利用自行编制的MATLAB高斯扩散程序得到爆炸危险区域的浓度曲线,分析环境风速对氢气扩散的影响,即风速越大,危险区域越向泄漏口收缩;利用CFD软件Fluent建立加氢站氢气泄漏全场景二维模型,模拟结果表明,无风情况下,氢气水平和垂直扩散速度很快,容易富集并形成爆炸气团,而在风速10 m/s情况下,泄漏氮气被带动、吹散和稀释,难以富集,爆炸区域仅限于泄漏点附近.环境风不利于氢气稳定扩散,对安全有利.

  19. 某涉氨制冷企业液氨储罐泄漏事故的后果分析%Analysis of An Ammonia Refrigeration Enterprise Involved in Liquid Ammonia Storage Tank Leakage Accident Consequence

    Institute of Scientific and Technical Information of China (English)

    周峰

    2014-01-01

    以某涉氨制冷企业液氨储罐为例,选用蒸气云爆炸、沸腾液体扩展蒸气爆炸和中毒模型对液氨储罐泄漏事故进行后果分析,定量地得出各类伤害半径,为企业制定应急救援预案和政府进行安全监管提供科学依据。%Taking the liquid ammonia tank of a refrigeration enterprise as example, the consequences of liquid ammonia storage tank leakage accident were analyzed using vapor cloud explosion model, boiling liq-uid expanding vapor explosion model and poisoning model. The various damage radiuses were calculated. It provided the scientific basis for enterprises to formulate emergency rescue plans and for safety supervision of government.

  20. Assessment of uncertainties in early off-site consequences from nuclear reactor accidents

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K.; Cazzoli, E.G. (Brookhaven National Lab., Dept. of Nuclear Energy, Upton, NY (US)); Khatib-Rahbar, M. (Energy Research, Inc., Rockville, MD (US))

    1990-04-01

    A simplified approach has been developed to calculate uncertainties in early off-site consequences from nuclear reactor accidents. The consequence model (SMART) is based on a solution procedure that uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarketing and detailed sensitivity and uncertainty analyses using SMART are presented.

  1. Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

    Energy Technology Data Exchange (ETDEWEB)

    White, J.E.; Roussin, R.W.; Gilpin, H.

    1988-12-01

    A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.

  2. Analysis on the `Thermite` reaction consequences in accidents involving research reactors using plate-type fuel; Analisis sobre las concequencias de la reaccion `Termita` en caso de accidentes en reactores de investigacion que utilizan combustible tipo placa

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L.; Bruno, Hernan R.; Camacho, Esteban F.; Cincotta, Daniel O.; Yorio, Daniel [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Constituyentes

    1999-11-01

    The mixture of Al-U{sub 3} O{sub 8} is not in a state of chemical equilibrium, and at temperatures of between 850 deg C and 1000 deg C, it reacts exo thermally. This is known, in corresponding bibliography as a `Thermite reaction. This mixture is used in the manufacturing of the plate-type fuel used in research reactors. It has been pointed out that the release of energy caused by this type of reactions might represent a risk in case of accidents in this type of reactor. Conclusions, in general, tend to indicate that no such risk exists, although no concrete assurance is given that this is the case, and this fact, therefore, leaves room for doubt. The objective of this paper is to provide an in-depth study of what happens to a fuel plate when it is subjected to thermite reaction. We will, furthermore, analyze the consequences of the release of energy generated by this type of reaction within the core of the reactor, clearly defining the problem for this type of fuel and this kind of reactor. (author) 3 refs., 9 figs., 1 tab.

  3. [Hanggliding accidents. Distribution of injuries and accident analysis].

    Science.gov (United States)

    Ballmer, F T; Jakob, R P

    1989-12-01

    Paragliding--a relatively new sport to Switzerland--brought 23 patients with 48 injuries (38% lower limb and 29% spinal) within a period of 8 months to the Inselspital University hospital in Berne. The aim of the study in characterizing these injuries is to formulate some guidelines towards prevention. With over 90% of accidents occurring at either take off or landing, emphasis on better training for the beginner is proposed with strict guidelines for the more experienced pilot flying in unfavourable conditions.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Palmrose, D E; Yang, J M

    2007-05-10

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

  6. NASA Accident Precursor Analysis Handbook, Version 1.0

    Science.gov (United States)

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

    2011-01-01

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

  7. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

  8. Energy Analysis of Road Accidents Based on Close-Range Photogrammetry

    Directory of Open Access Journals (Sweden)

    Alejandro Morales

    2015-11-01

    Full Text Available This paper presents an efficient and low-cost approach for energy analysis of road accidents using images obtained using consumer-grade digital cameras and smartphones. The developed method could be used by security forces in order to improve the qualitative and quantitative analysis of traffic accidents. This role of the security forces is crucial to settle arguments; consequently, the remote and non-invasive collection of accident related data before the scene is modified proves to be essential. These data, taken in situ, are the basis to perform the necessary calculations, basically the energy analysis of the road accident, for the corresponding expert reports and the reconstruction of the accident itself, especially in those accidents with important damages and consequences. Therefore, the method presented in this paper provides the security forces with an accurate, three-dimensional, and scaled reconstruction of a road accident, so that it may be considered as a support tool for the energy analysis. This method has been validated and tested with a real crash scene simulated by the local police in the Academy of Public Safety of Extremadura, Spain.

  9. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-23

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

  11. Hindsight Bias in Cause Analysis of Accident

    Institute of Scientific and Technical Information of China (English)

    Atsuo Murata; Yasunari Matsushita

    2014-01-01

    It is suggested that hindsight becomes an obstacle to the objective investigation of an accident, and that the proper countermeasures for the prevention of such an accident is impossible if we view the accident with hindsight. Therefore, it is important for organizational managers to prevent hindsight from occurring so that hindsight does not hinder objective and proper measures to be taken and this does not lead to a serious accident. In this study, a basic phenomenon potentially related to accidents, that is, hindsight was taken up, and an attempt was made to explore the phenomenon in order to get basically insights into the prevention of accidents caused by such a cognitive bias.

  12. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    CERN Document Server

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  13. Calculation notes that support accident scenario and consequence of the evaporator dump

    Energy Technology Data Exchange (ETDEWEB)

    Crowe, R.D., Westinghouse Hanford

    1996-09-09

    The purpose of this calculation note is to provide the basis for evaporator dump consequence for the Tank Farm Safety Analysis Report (FSAR). Evaporator Dump scenario is developed and details and description of the analysis methods are provided.

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

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    1999-10-20

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

  15. Development of MAAP5.0.3 Spent Fuel Pool Model for Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    After the Fukushima accident, the severe accident phenomena in the Spent Fuel Pool (SFP) have been the great issues in the nuclear industry. Generally, during full power operation status, the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident that is the say, the melting of fuel and fuel rack. In addition to this, the SFP of the PWR is not isolated within the containment like the SFP of the old BWR plant, there are so many possible measures to prevent and mitigate severe accidents in the SFP. On the other hand, in the low power shutdown status (fuel refueling), all the core is transferred into the SFP during the refueling period. At this period, if some accidents happen such as the loss of SFP cooling and the failure of SFP integrity then the accidents may be developed into severe accident because the decay heat is high enough. So, the analysis of severe accidents in the SFP during low power shutdown state is greatly affected to the establishment of the major strategies in the severe accident management guideline (SAMG). However, the status of the domestic technical background for those analyses is very weak. it is known that the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident qualitatively. However, there are some possibilities that can cause the severe accidents in the SFP if the loss of SFP cooling and integrity happens simultaneously. The severe accident phenomena in SFP themselves are not much different from those in the containment. However, since the structure of SFP cannot be isolated during the accidents like the containment, the consequence can be extremely significant. So, in terms of the establishment of the severe accident management strategy, it is necessary that the quantitative analysis for the severe accident progression in the SFP should be performed. In this study, the general behavior which can be appeared during the severe accidents in the SFP was analyzed using the

  16. INFLUENCE OF ANTIHYPERTENSIVE THERAPY ON PSYCHOLOGICAL STATUS OF CHERNOBYL NUCLEAR POWER PLANT ACCIDENT CONSEQUENCES LIQUIDATORS

    Directory of Open Access Journals (Sweden)

    E. M. Manoshkina

    2006-01-01

    Full Text Available Aim. To study psychological status and influence of antihypertensive therapy (AHT on it in Chernobyl nuclear power plant (NPP accident consequences liquidators, who suffer arterial hyper-tension (AH, with controlled treatment compared to the standard treatment in out-patient clinic. Material and methods. 81 liquidators with AH (all men were included into open compara-tive randomized study. Study duration was 12 months. Patients were randomized into main group (MG and control group (CG. Patients of MG received strictly regulated stepped AHT based on ACE inhibitor spirapril 6 mg daily (Quadropril®, Pliva-AVD, hypothiazide was added if necessary (12.5-25 mg daily and afterwards – atenolol (12.5-100 mg daily. In CG AHT and its correction was set by physician in polyclinic. Brief multifactor questionnaire for personality analysis was used to study psychological status. Results. 57 patients completed the study, 28 in MG and 29 in CG. In MG target blood pres-sure (BP levels were reached in 22 (78.6% patients, in CG – in 11 (38% patients (p<0.01. The main feature of psychological status of liquidators with AH was hypochondriac, depressive and anxious disorders. Controlled AHT made it possible to reach improvement in psychological status, i.e. growth of optimism and activity of patients, more often, than standard treatment in out-patient clinics. Increase in number of patients with pronounced anxious changes was observed in CG. Effi-ciency of AHT in liquidators with AH is connected with severity of depressive disturbances: in subgroups with inefficient treatment patients had the highest level of depression. In liquidators with AH, possessing neurotic disturbances, spirapril was efficient both as monotherapy, and in combina-tion with diuretic hydrochlorothiazide and beta-blocker atenolol. Conclusion. Controlled AHT in liquidators with AH has advantages over standard treatment in out-patient clinic and results in more frequent target BP level

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

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    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).

  18. Analysis of traffic accidents in children

    Directory of Open Access Journals (Sweden)

    Pavlekić Snežana

    2006-01-01

    Full Text Available Introduction: Violent health damages of different origin (accidents, murders, suicides in children and youth are one of the main causes of death and disabilities in this group of population in most countries. Objective: Objective of our paper was to analyze all related factors of traffic accidents involving children and to propose adequate measures of their prevention. Method: The analysis of fatal traffic accidents of children and youth aged to 18 years on the territory of Belgrade, within the period from 1998 to 2002. Results: In relation to other forms of violent death, the traffic mortality rate in children and youth holds the leading position, accounting for 56.9% with pedestrians as the most frequent category (57.4%. The most frequent age was between 7 and 9 years (46.8% and the boys were more frequently injured than the girls. It was established that the majority of children (51.9% was either running across the street outside the pedestrian/ zebra crossings or they were carelessly running out in the street, especially in April, July, August and September. More than a half of them (55.5%, predominantly school children, were injured by the end of working week, on Thursday and Friday. Conclusion: Results of our research have shown that the traffic education of children in our region is inadequate. Due to the abovementioned, it is primarily necessary to establish long-term and permanent education of this category of population. In addition, some public investments in the City infrastructure will be required in order to reduce the risk of traffic injuries in children.

  19. Development of the Severe Accident Analysis DB for the Severe Accident Management Expert System (I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    This report contains analysis methodologies and calculation results of 5 initiating events of the severe accident analysis database system. The Ulchin 3,4 NPP has been selected as reference plants. Based on the probabilistic safety analysis of the corresponding plant, 54 accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the Large loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data to develop the database system

  20. A 25 year retrospective review of the psychological consequences of the Chernobyl accident.

    Science.gov (United States)

    Bromet, E J; Havenaar, J M; Guey, L T

    2011-05-01

    The Chernobyl Forum Report from the 20th anniversary of the Chernobyl nuclear power plant disaster concluded that mental health effects were the most significant public health consequence of the accident. This paper provides an updated review of research on the psychological impact of the accident during the 25 year period since the catastrophe began. First responders and clean-up workers had the greatest exposure to radiation. Recent studies show that their rates of depression and post-traumatic stress disorder remain elevated two decades later. Very young children and those in utero who lived near the plant when it exploded or in severely contaminated areas have been the subject of considerable research, but the findings are inconsistent. Recent studies of prenatally exposed children conducted in Kiev, Norway and Finland point to specific neuropsychological and psychological impairments associated with radiation exposure, whereas other studies found no significant cognitive or mental health effects in exposed children grown up. General population studies report increased rates of poor self-rated health as well as clinical and subclinical depression, anxiety, and post-traumatic stress disorder. Mothers of young children exposed to the disaster remain a high-risk group for these conditions, primarily due to lingering worries about the adverse health effects on their families. Thus, long-term mental health consequences continue to be a concern. The unmet need for mental health care in affected regions remains an important public health challenge 25 years later. Future research is needed that combines physical and mental health outcome measures to complete the clinical picture.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  2. Exploratory analysis of Spanish energetic mining accidents.

    Science.gov (United States)

    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.

  3. Radiological consequence assessments of degraded core accident scenarios derived from a generic Level 2 PSA of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu; Ishikawa, Jun; Tomita, Kenichi; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-12-01

    The radiological consequence assessments have been made of postulated core damage accidents with source terms derived from a generic Level 2 PSA of a BWR carried out by the Japan Atomic Energy Research Institute (JAERI). The source terms used were for the five core damage accident sequences with the drywell and wetwell failure cases, the release control case by venting of the containment and the accident termination case by the containment spray. The radiological consequences have been assessed for individual dose, collective dose, individual risk of early health effects and individual risk of late health effects by a probabilistic accident consequence assessment code, OSCAAR developed in JAERI. Following conclusions were obtained for the assumed source terms. In case of the over pressure failures of the primary containment vessel, the early fatalities can be mitigated through the implementation of early countermeasures, and the late cancer fatalities remains small. For the release control and accident termination cases, the individual and collective doses to the public can be reduced without any countermeasures due to the release reduction of the volatile radionuclides such as iodine and cesium. (author)

  4. Source term analysis for a nuclear submarine accident

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Hugron, J.J.M.R. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    1999-07-01

    A source term analysis has been conducted to determine the activity release into the environment as a result of a large-break loss-of-coolant accident aboard a visiting nuclear-powered submarine to a Canadian port. This best-estimate analysis considers the fractional release from the core, and fission product transport in the primary heat transport system, primary containment (i.e. reactor compartment) and submarine hull. Physical removal mechanisms such as vapour and aerosol deposition are treated in the calculation. Since a thermalhydraulic analysis indicated that the integrity of the reactor compartment is maintained, release from the reactor compartment will only occur by leakage; however, it is conservatively assumed that the secondary containment is not isolated for a 24-h period where release occurs through an open hatch in the submarine hull. Consequently, during this period, the activity release into the atmosphere is estimated as 4.6 TBq, leading to a maximum individual dose equivalent of 0.5 mSv at 800 metres from the berthing location. This activity release is comparable to that obtained in the BEREX TSA study (for a similar accident scenario) but is four orders of magnitude less than that reported in the earlier Davis study where, unrealistically, no credit had been taken for the containment system or for any physical removal processes. (author)

  5. A COMPARISON OF SOME STATISTICAL TECHNIQUES FOR ROAD ACCIDENT ANALYSIS

    NARCIS (Netherlands)

    OPPE, S INST ROAD SAFETY RES, SWOV

    1992-01-01

    At the TRRL/SWOV Workshop on Accident Analysis Methodology, heldin Amsterdam in 1988, the need to establish a methodology for the analysis of road accidents was firmly stated by all participants. Data from different countries cannot be compared because there is no agreement on research methodology,

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

  7. Have the consequences of reactor accidents for the population been well assessed? Six questions to the experts in the field

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, Peter

    2016-07-15

    Six questions to the experts in the field are posed: (1) Why is the assessment of accident consequences not separated in long-term and peak exposure? (2) Why is the exposure due to I-131 seen critical mainly in regard to the thyroid? (3) Do you have any reliable relations of health risk versus peak exposure? (4) Why do you not abolish the LNT assumption and replace it with a threshold model? (5) Why do you include indirect, psycho-somatic effects in assessing the consequences of reactor accidents when this is not customary with accidents with often more casualties? (6) How can the number of Chernobyl-assigned thyroid cancers have risen from some 600 about to some 4,000 today, when the latency period is in the range of 4 to 5 years?.

  8. Chernobyl Nuclear Reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  9. Analysis of local subassembly accident in KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Jeong, Kwan Seong; Hahn, Do Hee

    2000-10-01

    Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage.

  10. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

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

  11. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2016-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

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

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

    Science.gov (United States)

    Groen, Frank; Lutomski, Michael

    2010-01-01

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

  14. A dynamic food-chain model and program for predicting the consequences of nuclear accident

    Institute of Scientific and Technical Information of China (English)

    1998-01-01

    A dynamic food-chain model and program, DYFOM-95, forpredicting the radiological consequences of nuclear accident hasbeen developed, which is not only suitable to the West food-chainbut also to Chinese food chain. The following processes, caused byaccident release which will make an impact on radionuclideconcentration in the edible parts of vegetable are considered: dryand wet deposition interception and initial retention,translocation, percolation, root uptake and tillage. Activityintake rate of animals, effects of processing and activity intakeof human through ingestion pathway are also considered incalculations. The effects of leaf area index LAI of vegetable areconsidered in dry deposition model. A method for calculating thecontribution of rain with different period and different intensityto total wet deposition is established. The program contains 1 maincode and 5 sub-codes to calculate dry and wet deposition on surfaceof vegetable and soil, translocation of nuclides in vegetable,nuclide concentration in the edible parts of vegetable and inanimal products and activity intake of human and so on.

  15. Consequences of the Chernobyl accident for the natural and human environments

    Energy Technology Data Exchange (ETDEWEB)

    Dreicer, M. [Lawrence Livermore National Lab., CA (United States); Aarkog, A. [Risoe National Lab., Roskilde (Denmark); Alexakhin, R. [Russian Inst. of Agricultural Radiology and Agroecology (Russian Federation); Anspaugh, L. [Lawrence Livermore National Lab., CA (United States); Arkhipov, N.P. [Scientific and Technical Centre of the RIA `Pripyat` (Ukraine); Johansson, K.-J. [Swedish Univ. of Agricultural Sciences, Uppsala (Sweden)

    1996-07-01

    In the ten years since the Chernobyl accident, an enormous amount of work has been done to assess the consequences to the natural and human environment. Although it is difficult to summarize such a large and varied field, some general conclusions can be drawn. This background paper includes the main findings concerning the direct impacts of radiation on the flora and fauna; the general advances of knowledge in the cycling of radionuclides in natural, seminatural and agricultural environments; some evaluation of countermeasures that were used; and a summary of the human radiation doses resulting from the environmental contamination. although open questions still remain, it can be concluded that: (1) at high radiation levels, the natural environment has shown short term impacts but any significant long term impacts remain to be seen; (2) effective countermeasures can be taken to reduce the transfer of contamination from the environment to humans but these are highly site specific and must be evaluated in terms of practicality as well as population does reduction; (3) the majority of the doses have already been received by the human population. If agricultural countermeasures are appropriately taken, the main source of future doses will be the gathering of food and recreational activities in natural and seminatural ecosystems.

  16. Fukushima accident: the consequences in Japan, France and in Japan; Accident de Fukushima: les repercusions au Japon, en France et dans le Japon

    Energy Technology Data Exchange (ETDEWEB)

    Foucher, N.; Sorin, F.

    2011-03-15

    This document begins with a description of the Fukushima accident, the second article reviews the main consequences in Japan of the accident: setting of a forbidden zone around the plant, restriction of the exports of food products, or the shutdown of the Hamaoka plant. The third article is the reporting of an interview of L. Oursel, deputy general director of the Areva group, this interview deals mainly with the safety standard of the EPR and with the issue of passive safety systems. The last part of the document is dedicated to the consequences in France (null sanitary impact, cooperation between Areva, EdF, CEA and the Japanese plant operator Tepco...) and in the rest of the world: the organization of resistance tests in the nuclear power plants operating in the European Union, the decision about the agreement of EPR and AP1000 reactor has been delayed in United-Kingdom, acceleration of the German program for abandoning nuclear energy, Italy suspends its nuclear program, China orders a general overhaul of the safety standard of its nuclear power plants, Poland and Romania reaffirm their trust in nuclear energy, France wishes a 'mechanism' allowing a quick international intervention in case of major nuclear accident, Russia proposes measures to improve nuclear safety. (A.C.)

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

    Science.gov (United States)

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

    2010-01-01

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

  18. Development of economic consequence methodology for process risk analysis.

    Science.gov (United States)

    Zadakbar, Omid; Khan, Faisal; Imtiaz, Syed

    2015-04-01

    A comprehensive methodology for economic consequence analysis with appropriate models for risk analysis of process systems is proposed. This methodology uses loss functions to relate process deviations in a given scenario to economic losses. It consists of four steps: definition of a scenario, identification of losses, quantification of losses, and integration of losses. In this methodology, the process deviations that contribute to a given accident scenario are identified and mapped to assess potential consequences. Losses are assessed with an appropriate loss function (revised Taguchi, modified inverted normal) for each type of loss. The total loss is quantified by integrating different loss functions. The proposed methodology has been examined on two industrial case studies. Implementation of this new economic consequence methodology in quantitative risk assessment will provide better understanding and quantification of risk. This will improve design, decision making, and risk management strategies.

  19. [Severe parachuting accident. Analysis of 122 cases].

    Science.gov (United States)

    Krauss, U; Mischkowsky, T

    1993-06-01

    Based on a population of 122 severely injured patients the causes of paragliding accidents and the patterns of injury are analyzed. A questionnaire is used to establish a sport-specific profile for the paragliding pilot. The lower limbs (55.7%) and the lower parts of the spine (45.9%) are the most frequently injured parts of the body. There is a high risk of multiple injuries after a single accident because of the tremendous axial power. The standard of equipment is good in over 90% of the cases. Insufficient training and failure to take account of geographical and meteorological conditions are the main determinants of accidents sustained by paragliders, most of whom are young. Nevertheless, 80% of our patients want to continue paragliding. Finally some advice is given on how to prevent paragliding accidents and injuries.

  20. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  1. The relationship between alcohol and traffic accidents an actuarial analysis.

    NARCIS (Netherlands)

    Nelker, G.

    1970-01-01

    This analysis deals with the accident chances of the excessive drinkers, moderate drinkers and non-drinkers. The Swedish motor insurance results for total abstainers are compared to the average policyholders.

  2. MELCOR DB Construction for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y. M.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The Korea Atomic Energy Research Institute (KAERI) has been constructing a severe accident analysis database (DB) under a National Nuclear R and D Program. In particular, an MAAP (commercial code being widely used in industries for integrated severe accident analysis) DB for many scenarios including a station blackout (SBO) has been completed. This paper shows the MELCOR DB construction process with examples of SBO scenarios, and the results will be used for a comparison with the MAAP DB

  3. Calculation notes that support accident scenario and consequence development for the leak from a railcar/tank trailer at the 204-ar waste unloading facility

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

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

  4. Effect of the Duration Time of a Nuclear Accident on Radiological Health Consequences

    Directory of Open Access Journals (Sweden)

    Hyojoon Jeong

    2014-03-01

    Full Text Available This study aimed to quantify the effect of duration time of a nuclear accident on the radiation dose of a densely populated area and the resulting acute health effects. In the case of nuclear accidents, the total emissions of radioactive materials can be classified into several categories. Therefore, the release information is very important for the assessment of risk to the public. We confirmed that when the duration time of the emissions are prolonged to 7 hours, the concentrations of radioactive substances in the ambient air are reduced by 50% compared to that when the duration time of emission is one hour. This means that the risk evaluation using only the first wind direction of an accident is very conservative, so it has to be used as a screening level for the risk assessment. Furthermore, it is judged that the proper control of the emission time of a nuclear accident can minimize the health effects on residents.

  5. RADIATION HYGIENIC CONSEQUENCES OF THE ACCIDENT AT THE CHERNOBYL NPP AND THE TASKS OF THEIR MINIMIZATION

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2009-01-01

    Full Text Available The paper presents data on the role and results of activities of Rospotrebnadzor bodies and institutions in the field of ensuring population radiation protection during various periods since accident at the Chernobyl NPP. Radiation hygienic characterization of territories affected by radioactive contamination from the accident, population exposure dose range, issues of ensuring radiological well-being of population and ways of their solution are being presented in the paper.

  6. Analysis of Credible Accidents for Argonaut Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-04-01

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

  7. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-17

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

  9. Estimation of the Radiological Consequences of Fukushima Dai-ichi Nuclear Power Plant Accident using MACCS2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Min, Byung-Il; Park, Kihyun; Yang, Byung-Mo; Suh, Kyung-suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Three of them have undergone fuel melting and hydrogen explosions. A significant amount of radioactive material was released into the atmosphere from FDNPP and dispersed all over the world. In this study, we assessed the offsite consequences of Fukushima disaster in the region within a 30-km radius of FDNPP using the MELCOR Accident Consequence Code Systems 2(MACCS2) code, which is the Nuclear Regulatory Commission's (NRC's) code. The reflection of the realistic regional characteristics, such as long-term meteorological data, site- and population-specific data, and radiation safety regulatory, is essential to accurately analyze the off-site consequences. The assessment that reflects regional characteristics would contribute to identify main causes of exposure doses and to find the effective countermeasures for minimizing the accidental off-site consequences.

  10. Emergency Responses and Health Consequences after the Fukushima Accident; Evacuation and Relocation.

    Science.gov (United States)

    Hasegawa, A; Ohira, T; Maeda, M; Yasumura, S; Tanigawa, K

    2016-04-01

    The Fukushima accident was a compounding disaster following the strong earthquake and huge tsunami. The direct health effects of radiation were relatively well controlled considering the severity of the accident, not only among emergency workers but also residents. Other serious health issues include deaths during evacuation, collapse of the radiation emergency medical system, increased mortality among displaced elderly people and public healthcare issues in Fukushima residents. The Fukushima mental health and lifestyle survey disclosed that the Fukushima accident caused severe psychological distress in the residents from evacuation zones. In addition to psychiatric and mental health problems, there are lifestyle-related problems such as an increase proportion of those overweight, an increased prevalence of hypertension, diabetes mellitus and dyslipidaemia and changes in health-related behaviours among evacuees; all of which may lead to an increased cardiovascular disease risk in the future. The effects of a major nuclear accident on societies are diverse and enduring. The countermeasures should include disaster management, long-term general public health services, mental and psychological care, behavioural and societal support, in addition to efforts to mitigate the health effects attributable to radiation.

  11. Evaluation of sanitary consequences of Chernobylsk accident in France. Epidemiological surveillance plan, state of knowledge, risks evaluation and perspectives; Evaluation des consequences sanitaires de l'accident de Tchernobyl en France. Dispositif de surveillance epidemiologique, etat des connaissances, evaluation des risques et perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Verger, P.; Cherie-Challine, L

    2000-12-15

    This report jointly written by IPSN and InVS, reviews the sanitary consequences in France of the Chernobyl accident, which occurred in 1986. The first point is dedicated to a short presentation of the knowledge relative to the sanitary consequences of the Chernobyl accident in the high contaminated countries and to the risk factors of the thyroid cancer. Secondly, this report describes the main systems of epidemiological surveillance of health implemented in France in 1986 and in 1999, as well as the data of the incidence and mortality of thyroid cancer observed in France since 1975. In addition, this report presents an analysis of the risk of thyroid cancer related to radioactive contamination in France, for young people of less than 15 years of age who where living in 1986 in the highest contaminated areas of France (Eastern territories). For this purpose, the theoretical number of thyroid cancers in excess is evaluated for this population, on the basis of different available risk model. Finally starting from the results of risk assessment, there is a discussion about the relevance and the feasibility of different epidemiological methods in view of answering the questions related to the sanitary consequences of the Chernobyl accident. In conclusion, this report recommends to reinforce the surveillance of thyroid cancer in France. (author)

  12. Analysis of FY79 Army Aircraft Accidents.

    Science.gov (United States)

    1980-04-01

    materiel failure/ malfunction 1. A accident. taskerrr (E) s Jb prforanc whch evited The acronym for the 3W approach to the Investigation, fro tht rquied y... lulli , It t , 1AZ" l iil~ i ..-... lI :.].,;.’,, - ,- S !5U" i_ ’ , o ’o , "% ", . #.,""" ".’ - "-’.’’’’-. "" " "’ ."- " " " II I S.7 44’ * U 11 Imem <I

  13. Criticality accident detector coverage analysis using the Monte Carlo Method

    Energy Technology Data Exchange (ETDEWEB)

    Zino, J.F.; Okafor, K.C.

    1993-12-31

    As a result of the need for a more accurate computational methodology, the Los Alamos developed Monte Carlo code MCNP is used to show the implementation of a more advanced and accurate methodology in criticality accident detector analysis. This paper will detail the application of MCNP for the analysis of the areas of coverage of a criticality accident alarm detector located inside a concrete storage vault at the Savannah River Site. The paper will discuss; (1) the generation of fixed-source representations of various criticality fission sources (for spherical geometries); (2) the normalization of these sources to the ``minimum criticality of concern`` as defined by ANS 8.3; (3) the optimization process used to determine which source produces the lowest total detector response for a given set of conditions; and (4) the use of this minimum source for the analysis of the areas of coverage of the criticality accident alarm detector.

  14. Stochastic Consequence Analysis for Waste Leaks

    Energy Technology Data Exchange (ETDEWEB)

    HEY, B.E.

    2000-05-31

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

  15. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

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

  16. RESPONSIBILITY OF PHYSICAL EDUCATION TEACHER: CONSEQUENCES OF THE LEGAL CLAIMS IN ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Roberto Silva Piñeiro

    2015-12-01

    Full Text Available Being physical education an area that collects some case law, and that the professionalization required studies specifically, a review of appeals and complaints concerning accidents in school physical education, including sessions inside and outside. It was studied the sense of judicial and administrative resolutions about school accidents in physical education in Spain between 1988-2012, and its effects on physical education professionals. Most opinions and judgments studied the claims were rejected for various reasons, among them the casuality and risk taking, although there are outstanding judgments, blaming the teacher for not being present in class and for not preventing situations. The administration usually paid, although in some cases the teacher also participates.

  17. Pedal cycling accidents--mechanisms and consequences. A study from northern Sweden.

    Science.gov (United States)

    Björnstig, U; Näslund, K

    1984-01-01

    During one year, 447 persons attended the University Hospital of Umeå (Sweden) because of bicycling accidents. The incidence was highest in children, falling with advancing age. The most common accident was falling off a bicycle on an uneven or slippery road. Collisions and objects interfering with the rear or front wheel were also common causes. A high percentage of the injuries involved the head, and one-third of these were major injuries. Almost one-fifth of the injured received in-patient care (average 6 days) and a similar number were paid sickness benefit (average 26.5 days). Costs for treatment and benefit were estimated as approx. 2200 Swedish kronor (SEK) per injured person (1 USD = 4:30 SEK, 1979, and 1984 = 8:20 SEK).

  18. Analysis of the 1957-58 Soviet nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Trabalka, J.R.; Eyman, L.D.; Auerbach, S.I.

    1979-12-01

    The occurrence of a Soviet accident in the winter of 1957-58, involving the atmospheric release of reprocessed fission wastes (cooling time approximately 1-2 yrs.), appears to have been confirmed, primarily by an analysis of the USSR radioecology literature. Due to the high population density in the affected region (Cheliabinsk Province in the highly industrialized Urals Region) and the reported level of /sup 90/Sr contamination , the event probably resulted in the evacuation and/or resettlement of the human population from a significant area (100-1000 km/sup 2/). The resulting contamination zone is estimated to have contained approximately 10/sup 6/ Ci of /sup 90/Sr (reference radionuclide); a relatively small fraction of the total may have been dispersed as an aerosol. Although a plausible explanation for the incident exists (i.e., use of now-obsolete waste storage-/sup 137/Cs isotope separation techniques), it is not yet possible, based on the limited information presently available, to completely dismiss this phenomenon as a purely historical event. It seems imperative that we have a complete explanation of the causes and consequences of this incident. Soviet experience gained in application of corrective measures would be invaluable to the rest of the world nuclear community.

  19. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    Energy Technology Data Exchange (ETDEWEB)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

  20. Detection and analysis of accident black spots with even small accident figures.

    NARCIS (Netherlands)

    Oppe, S.

    1982-01-01

    Accident black spots are usually defined as road locations with high accident potentials. In order to detect such hazardous locations we have to know the probability of an accident for a traffic situation of some kind, or the mean number of accidents for some unit of time. In almost all procedures

  1. MELCOR accident analysis for ARIES-ACT

    Energy Technology Data Exchange (ETDEWEB)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Galyean, W.J.; Brownson, D.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

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

  3. Advanced accident sequence precursor analysis level 2 models

    Energy Technology Data Exchange (ETDEWEB)

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L. [and others

    1996-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Klishin, G.S.; Seleznev, V.E.; Aleoshin, V.V. [RFNC-VNIIEF (Russian Federation)

    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

  5. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  6. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

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

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D C; Blond, R M

    1980-01-01

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

  8. Scram discharge volume break studies accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

    This paper is a summary of a report describing the predicted response of Unit 1 at the Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant to a hypothetical small break loss of coolant accident (SBLOCA) outside of containment. The accident studied would be initiated by a break in the scram discharge volume (SDV) piping when it is pressurized to full reactor vessel pressure as a normal consequence of a reactor scram. If the scram could be reset, the scram outlet valves would close to isolate the SDV and the piping break from the reactor vessel. However, reset is possible only if the conditions that caused the scram have cleared; it has been assumed in this study that the scram signal remains in effect over a long period of time.

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

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

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

  10. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Chung, D.Y.

    1999-07-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45.

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

    Science.gov (United States)

    Mayrhofer, Hannes; Quendler, Elisabeth; Boxberger, Josef

    2014-01-01

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

  12. Probability and consequences of severe reactor accidents. 60th year atw

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, Norman Carl [Massachusetts Institute of Technology (MIT), Cambridge, MA (United States). Dept. of Nuclear Engineering

    2015-06-15

    The study carried out on behalf of former USAEC (United States Atomic Energy Commission) led by Prof. Rasmussen and published in reworked form as WASH 1400 by the USNRC (United States Nuclear Regulatory Commission) in 1975, assessed in 3,300 pages the risks that can be deducted from severe accidents in nuclear power plants. The results, often quoted and criticised, were so far the most conclusive statements to this question. In his lecture at the reactor meeting in 1976, Prof. Rasmussen tried to trace back the conclusion of the results to the question: Is the use of larger nuclear power plants, in accordance to experiences and calculations so far, acceptable? His risk assessment, related to American power plants and cites, on behalf of the BMI is currently evaluated by the IRS together with the LRA on specific occurrences within the Federal Republic of Germany.

  13. The French-German initiative for Chernobyl: programme 3: Health consequences of the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Tirmarche, M. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Radiological Protection and Human Health Div. (DRPH), Radiobiology and Epidemiology Dept., 92 - Fontenay-aux-Roses (France); Kellerer, A.M. [Munchen Univ., Strahlenbiologisches Institut (Germany); Bazyka, D. [Chornobyl Center (CC), Kiev regoin (Ukraine)

    2006-07-01

    - Goals: The main objectives of the health programme are collection and validation of existing data on cancer and non cancer diseases in the most highly contaminated regions of Ukraine, Russia and Belarus, common scientific expertise on main health indicators and reliable dosimetry, and finally communication of the results to the scientific community and to the public. - General Tasks: 1- Comparison between high and low exposed regions, 2- Description of trends over time, 3- Consideration of specific age groups. This methodological approach is applied on Solid cancer incidence and leukaemia incidence in different regions in Ukraine, Belarus and Russia, With a special focus on thyroid cancer in young exposed ages. - Thyroid cancer: Those exposed in very young ages continue to express a relatively high excess of thyroid cancer even though they have now reached the age group 15-29. Those exposed as young adults show a small increase, at least partly due to better screening conditions - Leukemia: Description of leukemia trends for various age groups show no clear difference between exposed and unexposed regions when focusing on those exposed at very young ages. The rates of childhood leukemia before and after the accident show no evidence of any increase (oblasts in Belarus over 1982-1998). - Specific studies: Incidence of congenital malformations in Belarus; Infant mortality and morbidity in the most highly contaminated regions; Potential effects of prenatal irradiation on the brain as a result of the Chernobyl accident; Nutritional status of population living in regions with different levels of contamination; Dosimetry of Chernobyl clean-up workers; Radiological passports in contaminated settlements. - Congenital malformations: As a national register was existing since the 1980's and gives the possibility to compare trends before and after the accident, results of congenital malformations describe large results collected over Belarus, There is no evidence of a

  14. Radionuclides release possibility analysis of MSR at various accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    There are some accidents which go beyond our expectation such as Fukushima Daiichi nuclear disaster and amounts of radionuclides release to environment, so more effort and research are conducted to prevent it. MSR (Molten Salt Reactor) is one of GEN-IV reactor types, and its coolant and fuel are mixtures of molten salt. MSR has a schematic like figure 1 and it has different features with the solid fuel reactor, but most important and interesting feature of MSR is its many safety systems. For example, MSR has a large negative void coefficient. Even though power increases, the reactor slows down soon. Radionuclides release possibility of MSR was analyzed at various accident conditions including Chernobyl and Fukushima ones. The MSR was understood to prevent the severe accident by the negative reactivity coefficient and the absence of explosive material such as water at the Chernobyl disaster condition. It was expected to contain fuel salts in the reactor building and not to release radionuclides into environment even if the primary system could be ruptured or broken and fuel salts would be leaked at the Fukushima Daiichi nuclear disaster condition of earthquake and tsunami. The MSR, which would not lead to the severe accident and therefore prevents the fuel release to the environment at many expected scenarios, was thought to have priority in the aspect of accidents. A quantitative analysis and a further research are needed to evaluate the possibility of radionuclide release to the environment at the various accident conditions based on the simple comparison of the safety feature between MSR and solid fuel reactor.

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

    Science.gov (United States)

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

    2014-06-01

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

  16. A System Supporting the Analysis of Motorway Traffic Accidents

    Directory of Open Access Journals (Sweden)

    Davide Anghinolfi

    2015-12-01

    Full Text Available This work presents a business intelligence tool for monitoring traffic accidents on motorways and supporting decisions relevant to road safety. The system manages information on road characteristics, traffic accidents and traffic volumes and produces reports for monitoring the evolution of key performance indicators for road safety, supporting decisions on actions for risk mitigation and safety improvements for road users. The paper illustrates the different types of analyses performed by the system. Pattern based analysis is used to evaluate safety performance indicators for the road sections matching defined patterns. Two different road segmentation algorithms, used to identify the most critical road sections according to various severity indicators, are presented and discussed. Differential analysis compares the value of selected severity indicators before and after the implementation of an intervention on a road. Finally, a graphical user interface allows the accident locations to be visualized and accidents with specific characteristics to be highlighted. The system was evaluated on the data collected between 2009 and 2011 for the A15 motorway in Italy, connecting Parma to La Spezia.

  17. INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    D.A. Kalinich

    1999-09-27

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

  18. Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

    Energy Technology Data Exchange (ETDEWEB)

    Nalbandyan, A.; Ytre-Eide, M.A.; Thoerring, H.; Liland, A.; Bartnicki, J.; Balonov, M.

    2012-06-15

    The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

  19. An overview of current knowledge concerning the health and environmental consequences of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident.

    Science.gov (United States)

    Aliyu, Abubakar Sadiq; Evangeliou, Nikolaos; Mousseau, Timothy Alexander; Wu, Junwen; Ramli, Ahmad Termizi

    2015-12-01

    Since 2011, the scientific community has worked to identify the exact transport and deposition patterns of radionuclides released from the accident at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in Japan. Nevertheless, there still remain many unknowns concerning the health and environmental impacts of these radionuclides. The present paper reviews the current understanding of the FDNPP accident with respect to interactions of the released radionuclides with the environment and impacts on human and non-human biota. Here, we scrutinize existing literature and combine and interpret observations and modeling assessments derived after Fukushima. Finally, we discuss the behavior and applications of radionuclides that might be used as tracers of environmental processes. This review focuses on (137)Cs and (131)I releases derived from Fukushima. Published estimates suggest total release amounts of 12-36.7PBq of (137)Cs and 150-160PBq of (131)I. Maximum estimated human mortality due to the Fukushima nuclear accident is 10,000 (due to all causes) and the maximum estimates for lifetime cancer mortality and morbidity are 1500 and 1800, respectively. Studies of plants and animals in the forests of Fukushima have recorded a range of physiological, developmental, morphological, and behavioral consequences of exposure to radioactivity. Some of the effects observed in the exposed populations include the following: hematological aberrations in Fukushima monkeys; genetic, developmental and morphological aberrations in a butterfly; declines in abundances of birds, butterflies and cicadas; aberrant growth forms in trees; and morphological abnormalities in aphids. These findings are discussed from the perspective of conservation biology.

  20. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

  1. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-15

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR.

  2. COMPARATIVE ANALYSIS OF SOME HEALTH INDICATORS OF VARIOUS RADIATION ACCIDENTS LIQUIDATORS

    Directory of Open Access Journals (Sweden)

    V. M. Shubik

    2010-01-01

    Full Text Available The article presents the results of comparative investigation of morbidity and immunity of liquidators of radiation accidents occurred in South Urals (Kyshtym accident, at Chernobyl NPP and nuclear submarines (NS consequences. The most evident immunity and health changes were revealed for liquidators of Chernobyl NPP accident (ChNPP. Investigations of Kyshtym accident liquidators revealed long-term immunological losses. Comparison of health indicators of Chernobyl and nuclear submarine accident liquidators reveals the possibility of combined influence of radiation and stress on the immunity and health.

  3. Review of Severe Accident Phenomena in LWR and Related Severe Accident Analysis Codes

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

    Full Text Available Firstly, importance of severe accident provision is highlighted in view of Fukushima Daiichi accident. Then, extensive review of the past researches on severe accident phenomena in LWR is presented within this study. Various complexes, physicochemical and radiological phenomena take place during various stages of the severe accidents of Light Water Reactor (LWR plants. The review deals with progression of the severe accidents phenomena by dividing into core degradation phenomena in reactor vessel and post core melt phenomena in the containment. The development of various computer codes to analyze these severe accidents phenomena is also summarized in the review. Lastly, the need of international activity is stressed to assemble various severe accidents related knowledge systematically from research organs and compile them on the open knowledge base via the internet to be available worldwide.

  4. Radiological consequences of the Chernobyl reactor accident; Radiologische Folgen des Tschernobyl-Ungluecks

    Energy Technology Data Exchange (ETDEWEB)

    Jacob, P.

    1996-05-01

    Large areas of Belarus, Russia, and the Ukraine have been highly contaminated by the radioactive fallout from the reactor accident at Chernobyl. The most affected areas are around Chernobyl and east of Gomel in Belarus, where part of the radioactive fallout came down with rain. The article maps the radioactive contamination through cesium 137 and iodine 131, and summarizes the immediate action taken at the time, as well as long-term remedial action for decontamination of soils. Data are given on the radiation exposure of the population, in particular doses to the thyroid, and prognoses on the incidence of thyroid cancer. (VHE) [Deutsch] Durch den Reaktorunfall von Tschernobyl wurden groessere Flaechen von Belarus, Russland und der Ukraine stark radioaktiv kontaminiert. Besonders betroffen sind die Umgebung von Tschernobyl sowie die Gegend oestlich von Gomel (Belarus), wo die radioaktive Wolke teilweise ausregnete. Der Artikel beschreibt die Belastung mit Caesium 137 und Iod 131 sowie die ergriffenen Sofortmassnahmen und die langfristigen Massnahmen zur Dekontamination der betroffenen Boeden. Die Strahlenbelastung der Bevoelkerung, v.a. die Schilddruesendosen, werden beschrieben, fuer Schilddruesenkrebs werden Prognosen gegeben. (VHE)

  5. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    Energy Technology Data Exchange (ETDEWEB)

    Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.

  6. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

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

    2015-04-01

    . The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

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

    Science.gov (United States)

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

    2014-12-01

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

  8. Cassini Spacecraft Uncertainty Analysis Data and Methodology Review and Update/Volume 1: Updated Parameter Uncertainty Models for the Consequence Analysis

    Energy Technology Data Exchange (ETDEWEB)

    WHEELER, TIMOTHY A.; WYSS, GREGORY D.; HARPER, FREDERICK T.

    2000-11-01

    Uncertainty distributions for specific parameters of the Cassini General Purpose Heat Source Radioisotope Thermoelectric Generator (GPHS-RTG) Final Safety Analysis Report consequence risk analysis were revised and updated. The revisions and updates were done for all consequence parameters for which relevant information exists from the joint project on Probabilistic Accident Consequence Uncertainty Analysis by the United States Nuclear Regulatory Commission and the Commission of European Communities.

  9. The reactor accident in Fukushima Daiichi. The consequence of design deficiencies and inadequate safety engineering; Der Reaktorunfall in Fukushima Daiichi. Folge fehlerhafter Auslegung und unzureichender Sicherheitstechnik

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-03-15

    The reactor accident in Fukushima Daiichi is discussed in the frame of design deficiencies and inadequate safety engineering. The progress of the accident as consequence of the earthquake and the tsunami is described. The radiological situation for the public is supposed to be blow the dose limit of 20 mSv/year. The WHO and UNSCEAR (United Nations Scientific Committee on the Effects of Atomic radiation) did not observe acute radiation injuries. The Japanese authorities have classified the accident to 7 of the INES scale. The German Atomforum e.V. considers the safety engineering of German NPPs to be superior to the Japanese situation due to higher emergency energy supply, extensive measures to reduce the hydrogen accumulation and mitigating measures for the accident management. German NPPS are considered highly robust as the EU stress tests have shown.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

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

  11. Radiological Consequence Analysis of Annulus Ventilation System with Delayed Operation under Severe Accident of Nuclear Power Plants%核电厂严重事故下双层安全壳环形空间通风系统延迟投运的放射性后果影响分析

    Institute of Scientific and Technical Information of China (English)

    吴楠; 黄树明; 刘新建

    2016-01-01

    During severe accidents, forthenuclear power plants ( NPPs) with double⁃containment design, ifa sub⁃atmospheric pressure cannot be built or the accident filter cannot be activated when annulus ventila⁃tion system is failed to operate normally, the control function of fission⁃product releaseof dual containment would be weakened�With focus on the double⁃containment design adopted by most of the Gen-ⅢNPPs in the world, this paper firstly calculates the release amountof radioactive materialsinto environment under different delay scenarios of annulus ventilation systemoperation, with the consideration of the intact con⁃tainment and using NUREG-1465 source term� Then the Criteria for Limited Impact ( CLI) provided in European Utility Requirements ( EUR) areare applied to evaluateradiological consequence of severe acci⁃dent, and the relationship between the delay of annulus ventilation systemoperation and “large release” is analyzed� The results could beareference for the emergency response actions and radiological consequence estimation in the context of severe accidents.%核电厂严重事故工况下,对于具有双层安全壳设计的核电机组,若环形空间通风系统不能正常运转,无法形成负压或无法启动事故过滤器,双层安全壳对放射性物质释放的控制效果将被削弱。鉴于此,本文针对目前国际上多个第三代核电机组采用的双层安全壳设计,考虑安全壳完整并选用NUREG-1465源项作为严重事故源项,计算环形空间通风系统在不同延迟投运场景下放射性物质的环境释放量,同时采用“欧洲用户要求( EUR)”文件提出的有限影响准则对严重事故的放射性后果进行评价,分析环形空间通风系统的延迟投运同“大量释放”间的关系。研究结果可为严重事故下的应急响应行动及放射性后果评价提供参考。

  12. HTGR accident initiation and progression analysis status report. Volume V. AIPA fission product source terms

    Energy Technology Data Exchange (ETDEWEB)

    Alberstein, D.; Apperson, C.E. Jr.; Hanson, D.L.; Myers, B.F.; Pfeiffer, W.W.

    1976-02-01

    The primary objective of the Accident Initiation and Progression Analysis (AIPA) Program is to provide guidance for high-temperature gas-cooled reactor (HTGR) safety research and development. Among the parameters considered in estimating the uncertainties in site boundary doses are uncertainties in fission product source terms generated under normal operating conditions, i.e., fuel body inventories, circulating coolant activity, total plateout activity in the primary circuit, and plateout distributions. The volume presented documents the analyses of these source term uncertainties. The results are used for the detailed consequence evaluations, and they provide the basis for evaluation of fission products important for HTGR maintenance and shielding.

  13. Analysis on the nitrogen drilling accident of Well Qionglai 1 (II: Restoration of the accident process and lessons learned

    Directory of Open Access Journals (Sweden)

    Yingfeng Meng

    2015-12-01

    Full Text Available All the important events of the accident of nitrogen drilling of Well Qionglai 1 have been speculated and analyzed in the paper I. In this paper II, based on the investigating information, the well log data and some calculating and simulating results, according to the analysis method of the fault tree of safe engineering, the every possible compositions, their possibilities and time schedule of the events of the accident of Well Qionglai 1 have been analyzed, the implications of the logging data have been revealed, the process of the accident of Well Qionglai 1 has been restored. Some important understandings have been obtained: the objective causes of the accident is the rock burst and the induced events form rock burst, the subjective cause of the accident is that the blooie pipe could not bear the flow burden of the clasts from rock burst and was blocked by the clasts. The blocking of blooie pipe caused high pressure in wellhead, the high pressure made the blooie pipe burst, natural gas came out and flared fire. This paper also thinks that the rock burst in gas drilling in fractured tight sandstone gas zone is objective and not avoidable, but the accidents induced from rock burst can be avoidable by improving the performance of the blooie pipe, wellhead assemblies and drilling tool accessories aiming at the downhole rock burst.

  14. Analysis of Three Mile Island-Unit 2 accident

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-01

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

  15. Review of psychological consequences of nuclear accidents and empirical study on peoples reactions to radiation protection activities in an imagined situation.; Katsaus ydinonnettomuuksien psykologisiin seurauksiin sekae empiirinen tutkimus saeteilysuojelutoimenpiteiden vaikutuksista kaeyttaeytymiseen kuvitteelisessa tilanteessa

    Energy Technology Data Exchange (ETDEWEB)

    Haukkala, A.; Eraenen, L. [Helsinki Univ. (Finland). Dept. of Social Psychology

    1994-10-01

    The report consist of two parts: a review of studies on psychological consequences of nuclear and radiation accidents in population and an empirical study of peoples reactions to protection actions in an event of hypothetical accident. Review is based on research results from two nuclear reactor accidents (Three Mile Island 1979, Chernobyl 1986) and a radiation accident in Goiania, Brazil 1987. (53 refs, 2 figs.,7 tabs.).

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

    Science.gov (United States)

    Kara, Can; Akçit, Nuhcan

    2015-06-01

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

  17. Off-site consequences of radiological accidents: methods, costs and schedules for decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Tawil, J.J.; Bold, F.C.; Harrer, B.J.; Currie, J.W.

    1985-08-01

    This report documents a data base and a computer program for conducting a decontamination analysis of a large, radiologically contaminated area. The data base, which was compiled largely through interviews with knowledgeable persons both in the public and private sectors, consists of the costs, physical inputs, rates and contaminant removal efficiencies of a large number of decontamination procedures. The computer program utilizes this data base along with information specific to the contaminated site to provide detailed information that includes the least costly method for effectively decontaminating each surface at the site, various types of property losses associated with the contamination, the time at which each subarea within the site should be decontaminated to minimize these property losses, the quantity of various types of labor and equipment necessary to complete the decontamination, dose to radiation workers, the costs for surveying and monitoring activities, and the disposal costs associated with radiological waste generated during cleanup. The program and data base are demonstrated with a decontamination analysis of a hypothetical site. 39 refs., 24 figs., 155 tabs.

  18. Estimating the causes of traffic accidents using logistic regression and discriminant analysis.

    Science.gov (United States)

    Karacasu, Murat; Ergül, Barış; Altin Yavuz, Arzu

    2014-01-01

    Factors that affect traffic accidents have been analysed in various ways. In this study, we use the methods of logistic regression and discriminant analysis to determine the damages due to injury and non-injury accidents in the Eskisehir Province. Data were obtained from the accident reports of the General Directorate of Security in Eskisehir; 2552 traffic accidents between January and December 2009 were investigated regarding whether they resulted in injury. According to the results, the effects of traffic accidents were reflected in the variables. These results provide a wealth of information that may aid future measures toward the prevention of undesired results.

  19. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  20. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  1. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  2. Radiological consequence analysis with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fesak, G.M.; Breland, R.M.; Spadaro, J. [Dept. of Labor, Arlington, VA (United States)

    1996-12-31

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

  4. Improvement of severe accident analysis method for KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Cho, Song Won; Cho, Youn Soo [Korea Radiation Technology Institute Co., Taejon (Korea, Republic of)

    2002-03-15

    The objective of this study is preparation of MELCOR 1.8.5 input deck for KSNP and simulation of some major severe accidents. The contents of this project are preparation of MELCOR 1.8.5 base input deck for KSNP to understand severe accident phenomena and to assess severe accident strategy, preparation of 20 cell containment input deck to simulate the distribution of hydrogen and fission products in containment, simulation of some major severe accident scenarios such as TLOFW, SBO, SBLOCA, MBLOCA, and LBLOCA. The method for MELCOR 1.8.5 input deck preparation can be used to prepare the input deck for domestic PWRs and to simulate severe accident experiments such as ISP-46. Information gained from analyses of severe accidents may be helpful to set up the severe accident management strategy and to develop regulatory guidance.

  5. Administrative circular No.14 (Rev. 3) – Protection of members of the personnel against the financial consequences of illness, accident and incapacity for work

    CERN Multimedia

    2013-01-01

    Administrative Circular No. 14 (Rev. 3) entitled “Protection of members the personnel against the financial consequences of illness, accident and incapacity for work”, approved by the Director-General following discussion at the Standing Concertation Committee meeting of 19 April 2012 and entering into force on 1 January 2013, is available on the intranet site of the Human Resources Department.   This circular is applicable to all members of the personnel. It cancels and replaces Administrative Circular No. 14 (Rev. 2) entitled “Protection of members of the personnel against the financial consequences of illness, accident and disability” from July 2006. The circular was revised in order to improve the procedure before the Joint Advisory Rehabilitation and Disability Board (JARDB) and the management of long-term sick leave through a multidisciplinary approach launched upstream. The aim of this approach is to allow staff/fellows c...

  6. Comparison of Management Oversight and Risk Tree and Tripod-Beta in Excavation Accident Analysis

    Directory of Open Access Journals (Sweden)

    Mohamadfam

    2015-01-01

    Full Text Available Background Accident investigation programs are a necessary part in identification of risks and management of the business process. Objectives One of the most important features of such programs is the analysis technique for identifying the root causes of accidents in order to prevent their recurrences. Analytical Hierarchy Process (AHP was used to compare management oversight and risk tree (MORT with Tripod-Beta in order to determine the superior technique for analysis of fatal excavation accidents in construction industries. Materials and Methods MORT and Tripod-Beta techniques were used for analyzing two major accidents with three main steps. First, these techniques were applied to find out the causal factors of the accidents. Second, a number of criteria were developed for the comparison of the techniques and third, using AHP, the techniques were prioritized in terms of the criteria for choosing the superior one. Results The Tripod-Beta investigation showed 41 preconditions and 81 latent causes involved in the accidents. Additionally, 27 root causes of accidents were identified by the MORT analysis. Analytical hierarchy process (AHP investigation revealed that MORT had higher priorities only in two criteria than Tripod-Beta. Conclusions Our findings indicate that Tripod-Beta with a total priority of 0.664 is superior to MORT with the total priority of 0.33. It is recommended for future research to compare the available accident analysis techniques based on proper criteria to select the best for accident analysis.

  7. An analysis of accident data for franchised public buses in Hong Kong.

    Science.gov (United States)

    Evans, W A; Courtney, A J

    1985-10-01

    This paper analyses data on accidents involving franchised public buses operating in Hong Kong. The data were obtained from the Royal Hong Kong Police, the Hong Kong Government Transport Department, the two major franchised bus operators and international sources. The analysis includes an international comparison of accidents with emphasis on the situation in Hong Kong compared to urban areas in the United Kingdom. An attempt has been made to identify the characteristics of bus accidents; accident incidence has been related to time of day, day of the week, time of year, weather conditions, driver's age and experience, hours on duty and policy-reported cause. The results indicate that Hong Kong has a high accident rate compared to Japan, the U.K. and the U.S.A., with particularly high pedestrian involvement rates. Bus accidents peak at around 9:00 AM and 4:00 PM but the accident rate is high throughout the day. Monday and Saturday appear to have a higher than average accident rate. The variability of accident rate throughout the year does not seem to be significant and the accident rate does not appear to be influenced by weather conditions. Older, more experienced drivers generally have a safer driving record than their younger, less experienced colleagues. Accident occurrence is related to the time the driver has been on duty. The paper questions the reliability of police-reported accident causation data and suggests improvements in the design of the accident report form and in the training of police investigators. The relevance of the Hong Kong study for accident research in general is also discussed.

  8. Exploring the potential of data mining techniques for the analysis of accident patterns

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Bekhor, Shlomo; Galtzur, Ayelet

    2010-01-01

    Research in road safety faces major challenges: individuation of the most significant determinants of traffic accidents, recognition of the most recurrent accident patterns, and allocation of resources necessary to address the most relevant issues. This paper intends to comprehend which data mining...... and association rules) data mining techniques are implemented for the analysis of traffic accidents occurred in Israel between 2001 and 2004. Results show that descriptive techniques are useful to classify the large amount of analyzed accidents, even though introduce problems with respect to the clear...... importance of input and intermediate neurons, and the relative importance of hundreds of association rules. Further research should investigate whether limiting the analysis to fatal accidents would simplify the task of data mining techniques in recognizing accident patterns without the “noise” probably...

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

    Science.gov (United States)

    Kumar, R; Ghosh, A K

    2014-01-01

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

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

    Science.gov (United States)

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

    2009-01-01

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

  11. Radionuclide analysis on bamboos following the Fukushima nuclear accident.

    Directory of Open Access Journals (Sweden)

    Takumi Higaki

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

  12. Analysis on relation between safety input and accidents

    Institute of Scientific and Technical Information of China (English)

    YAO Qing-guo; ZHANG Xue-mu; LI Chun-hui

    2007-01-01

    The number of safety input directly determines the level of safety, and there exists dialectical and unified relations between safety input and accidents. Based on the field investigation and reliable data, this paper deeply studied the dialectical relationship between safety input and accidents, and acquired the conclusions. The security situation of the coal enterprises was related to the security input rate, being effected little by the security input scale, and build the relationship model between safety input and accidents on this basis, that is the accident model.

  13. Development of a post accident analysis model for KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Chang, W. P.; Ha, G. S.; Jeong, H. Y.; Kwon, Y. M.; Heo, S.; Lee, Y. B. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    An ultimate safety measure of the KALIMER depends on the inherent safety, which have the core maintain a negative reactivity during any accident periods. In order to secure the integrity of a fuel rod, the void reactivity feedback under sodium boiling must be analyzed. Even though the KALIMER design might not allow boiling at any circumstance, sodium boiling would be possible under HCDA (Hypothetical Core Disruptive Accident) initiating events which are represented by UTOP (Unprotected Transient Over Power), ULOF (Unprotected Loss Of Flow), ULOHS (Unprotected Loss Of Heat Sink), or sudden flow channel blockage, due to power excursion caused by the reactivity feedback. The slug and annular flow regimes tend to prevail for the boiling of a liquid-metal coolant such as sodium near the atmospheric pressure. In contrast, the bubbly flow is typical under a high pressure in light water reactors. This phenomenon difference brings to develop the present model, especially, at the onset of boiling. A few models had been developed for the sodium boiling analysis. The models such as those in the HOMSEP-2 and SAS series are classified into relatively detailed models. Both models are usually called a multiple-bubble slug ejection model. Some simpler models are also introduced to evade either parameter sensitivities or a mathematical complexity associated with those rigorous models. The present model based on the multiple-bubble slug ejection model. It allows a finite number (N) of bubbles, separated by liquid slugs, in a channel. Boiling occurs at a user specified superheat, and a generated vapor is modeled to fill the whole cross section of the coolant channel except for a static liquid film left on the cladding or/and structure surfaces. The model also assumes a vapor with one uniform pressure. The present analysis is focused on the behavior of early sodium boiling after ULOHS.

  14. Advanced accident sequence precursor analysis level 1 models

    Energy Technology Data Exchange (ETDEWEB)

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O. [Idaho National Engineering Lab., Idaho National Lab., Idaho Falls, ID (United States)

    1996-03-01

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

  15. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. 福岛核事故后果初步评价与思考%Consequences of the Fukushima Accident: A Preliminary Assessment and Discussion

    Institute of Scientific and Technical Information of China (English)

    张立国; 曹建主; 薛大知; 曲静原; 童节娟

    2012-01-01

    东日本的大地震引发的海啸造成日本福岛第一核电站发生严重核事故,引起了国内外社会广泛关注。对此次核事故放射性源项和事故所致后果进行了大致评价与预测。与后续事故发展情况相比较,本文评价工作从整体上把握了事故规模及其所致后果。%Tsunami due to the earthquake in East Japan Sea eventually leaded to a severe nuclear accident in Fukushima Dai-ichi nuclear power plant. This event immediately became the focus of the whole world. The work to roughly evaluate and predict the consequence of this nuclear accident is summarized in this paper and the work actually provides valuable information in predicting the scale and severity of the accident comparing to the published information on the accident thereafter.

  18. Cause Analysis of Wuhan Tianheng Building Pile Accident

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The geological condition and the original structure feature and foundation design of Wuhan Tianheng building are described. The accident appearance of pile foundation in the construction execution of work is illustrated. The generating source of this pile foundation accident is analyzed in great details.``

  19. Road Accident Analysis: A Case Study of Federal Route FT024 Yong Peng- Parit Sulong

    Directory of Open Access Journals (Sweden)

    Mohd Masirin Mohd Idrus

    2016-01-01

    Full Text Available Traffic accidents are considered as an unplanned and unfortunate event which is a serious concern to the community as well as the authority. An accident-counter measure can reduce the rate of road accidents by initially identifying critical locations. The total of road accidents along FT 024 between 2009 and 2012 is 907 cases. Road accidents during the same period range between 24 % and 26 % each year. These accidents killed 34 people and injuring another 101 people. This research aims to identify factors that may contribute to the cause of accidents and to study the effects of speed, volume and road geometry on road accidents. In this study, the locations labelled as km 1, km 2, km 4, km 8, km 5 and km 14 of Federal Route FT 024 Yong Peng – Parit Sulong were selected as the study-case sections based on Accident Point Weightage of ranking, in which each location has different road characteristics. Speed study was carried out at selected road sections to evaluate the influence of speed upon road accidents; and traffic volume count was conducted at the same selected road sections to determine the existing condition of the route. Besides, road geometry observations and measurements were also conducted at selected sections, they were also studied to evaluate influence of road designs upon road accidents. The extracted data were analyzed by using regression analysis on different variables to evaluate the relationship between accident Weightage point and other dependent variables that were considered to have considerable effects upon road accidents such as mean speed, volume, shoulder width, lane width and access point. P value below 0.05 was considered as statistically significant. After conducting data analysis, this study showed that the number of road accidents increases with the increment of speed and access point. On the other hand, volume has no strong relationship to road accidents which means that it may not have an effect on accidents

  20. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    To investigate the consequence of severe accidents in fusion reactor, a number of thermal hydraulics simulation codes were used (ECART, INTRA, ATHENA/RELAP and so on). MELCOR is chosen as the thermal hydraulics code to simulate the consequence of radioactive material release from accident in preliminary safety report. Capability of the simulation code for fusion reactor severe accident analysis is ability to simulate the hydraulic system in ITER and the transport phenomenon of radionuclides. MELCOR is a fully integrated code that models the accidents in Light Water Reactor (LWR). There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products (AP) of divertor or first-wall and activated corrosion products(ACP). In generic Site Safety Report (GSSR), the release guidelines for tritium and activation products are listed for normal operation, incidents, and accidents. And this guidelines presented in Table 1. Not only ITER, the KSTAR (Korea Superconducting Tokamak Advanced Research) is also developing fusion research reactor. The scale of facility is smaller than ITER but this small scale of facility offers the experimental flexibility to develop fusion technology. The major differences between KSTAR and ITER systems are presented in Table 2. Fusion source difference between KSTAR and ITER is D-D fusion reaction (Deuterium-Deuterium fusion reaction) and D-T fusion reaction (Deuterium-Tritium fusion reaction). This D-D fusion makes one tritium by 50 percent chance. The radioactivity of tritium is small to consider compared to radioactive materials in nuclear fission reactor. This reaction is presented in equation (1) In the present work, conservatively estimated tritium inventory amount in KSTAR is used with one of the most severe accident in ITER; Ex-vessel pipe break with Fusion Power Termination System (FPTS). The MELCOR KSTAR input is made by scaling down the ITER input deck. So, the detail system is not same

  1. Report on preliminary analysis of state of nuclear criticality accident at JCO at Tokaimura, Ibaraki, Japan (I)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, J.J.; Park, J.H.; Chang, J.H. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This preliminary report was prepared by the Special Task Force Team of KAERI in order to analysis status of nuclear criticality accident broken out at 10:35 September 30, 1999 at JCO nuclear conversion test facility located at Tokaimura, Ibaraki, Japan. The report was consisted of accident summary of cause of accident summary of cause of accident and response by relevant organizations, and preliminary technical analysis of radiation exposure of JCO workers, analysis of cause of accident, and accident assessment and preventive actions against criticality accident. It is expected that JCO accident, Japan's first nuclear criticality accident, would make significant effects to Japan nuclear policy and would be also a good example to Korea future actions to be taken in use and development of nuclear energy. 63 refs., 3 figs., 1 tab. (Author)

  2. An Evidential Reasoning-Based CREAM to Human Reliability Analysis in Maritime Accident Process.

    Science.gov (United States)

    Wu, Bing; Yan, Xinping; Wang, Yang; Soares, C Guedes

    2017-01-09

    This article proposes a modified cognitive reliability and error analysis method (CREAM) for estimating the human error probability in the maritime accident process on the basis of an evidential reasoning approach. This modified CREAM is developed to precisely quantify the linguistic variables of the common performance conditions and to overcome the problem of ignoring the uncertainty caused by incomplete information in the existing CREAM models. Moreover, this article views maritime accident development from the sequential perspective, where a scenario- and barrier-based framework is proposed to describe the maritime accident process. This evidential reasoning-based CREAM approach together with the proposed accident development framework are applied to human reliability analysis of a ship capsizing accident. It will facilitate subjective human reliability analysis in different engineering systems where uncertainty exists in practice.

  3. A study on the core analysis methodology for SMART CEA ejection accident-I

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Lee, Chung Chan; Kim, Kyo Yoon; Cho, Byung Oh

    1999-04-01

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs.

  4. An Evaluation of the Role that Traffic Culture Plays in Reducing Consequences of Accidents and Promoting Social Security and Order

    Directory of Open Access Journals (Sweden)

    Nasser Pourmoallem

    2013-01-01

    Full Text Available IntroductionAccidents and traffic security have become serious issues in our country, to the extent that most of the people and authorities are severely concerned about them. On the other hand, research shows that human factor has the most important role in the occurrence of accidents. According to the records, only %1 of all accidents in Iran are resulted from "vehicle malfunction" and “immunodeficiency of the roads”; while other events, directly or indirectly, are caused by human wrong operations. Analysis of various factors shows that the human factor is not an element, but is characterized by three axes: (1 drivers and pedestrians, (2 planning and legislation and (3 control factors. In this paper, approaches to develop transportation and traffic security through teaching traffic behaviors to road users are investigated in the framework of three scenarios. Also, the solutions for improving safety, traffic and transportation through culture and education have been investigated. Moreover, the behavior of road users has been studied in the form of these traffic scenarios. Material & MethodsIn scenario No. 1, the importance and the role of traffic culture and behavior in the development of traffic flow is investigated and the process of AHP is used to investigate the decision making processes about the improvement of traffic culture and behavior. In this scenario, the importance of culture together with the role that it plays in improving the safety and facilitative factors of transportation is evaluated. To this end, “improving traffic behavior and culture alongside of the improvement of transport safety and facilitation” is intended to be the assumed target. Therefore, all the factors and parameters effective on the improvement of traffic behavior and culture are the statistical variables in this study:•The training method (culture•The enforcement of traffic laws and regulations variable•The variable of social and psychological

  5. IRSN press briefing on the issue 'Fukushima, one year after': Situation of Fukushima Dai-ichi nuclear installations; Accident of the Fukushima Dai-ichi: briefing on the situation in February 2012; The Fukushima 1 accident one year after: assessment of environmental consequences in Japan; assessment of consequences of the Fukushima accident on the environment in Japan, one year after; Health consequences of the Fukushima Dai-ichi: situation briefing in February 2012; Point presse de l'IRSN sur le theme 'Fukushima, un an apres': Situation des installations nucleaires de Fukushima Dai-ichi; Accident survenu a la centrale de Fukushima Dai-Ichi Point de la situation en fevrier 2012; L'accident de Fukushima 1 an apres: bilan des consequences environnementales au Japon; bilan des consequences de l'accident de Fukushima sur l'environnement au Japon, un an apres l'accident; Les consequences sanitaires de l'accident de Fukushima Dai-ichi: point de situation en fevrier 2012

    Energy Technology Data Exchange (ETDEWEB)

    Charles, T.; Jourdain, Jean-Rene

    2012-02-28

    This document gathers reports and Power Point presentations (with maps, data tables and graphs) dealing with the Fukushima accident, one year after its occurrence. Different issues are addressed: the status of the nuclear installations, the situation of the installations and of the environment, assessments, measurements and investigations on the effects and consequences of the accident (radioactive releases and fallouts) on the ground and marine environment and on public health

  6. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

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

    Science.gov (United States)

    Battaglia, Massimo; Frey, Marco; Passetti, Emilio

    2014-01-01

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

  8. Analysis of accidents with organic material in health workers.

    Science.gov (United States)

    Vieira, Mariana; Padilha, Maria Itayra; Pinheiro, Regina Dal Castel

    2011-01-01

    This retrospective and descriptive study with a quantitative design aimed to evaluate occupational accidents with exposure to biological material, as well as the profile of workers, based on reporting forms sent to the Regional Reference Center of Occupational Health in Florianópolis/SC. Data collection was carried out through a survey of 118 reporting forms in 2007. Data were analyzed electronically. The occurrence of accidents was predominantly among nursing technicians, women and the mean age was 34.5 years. 73% of accidents involved percutaneous exposure, 78% had blood and fluid with blood, 44.91% resulted from invasive procedures. It was concluded that strategies to prevent the occurrence of accidents with biological material should include joint activities between workers and service management and should be directed at improving work conditions and organization.

  9. Traffic Accident Analysis Using Decision Trees and Neural Networks

    OpenAIRE

    Chong, Miao M.; Abraham, Ajith; Paprzycki, Marcin

    2004-01-01

    The costs of fatalities and injuries due to traffic accident have a great impact on society. This paper presents our research to model the severity of injury resulting from traffic accidents using artificial neural networks and decision trees. We have applied them to an actual data set obtained from the National Automotive Sampling System (NASS) General Estimates System (GES). Experiment results reveal that in all the cases the decision tree outperforms the neural network. Our research analys...

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

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    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

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

    Science.gov (United States)

    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.

  12. Neutronic analysis of LMFBRs during severe core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Tomlinson, E.T.

    1979-01-01

    A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-23

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

  14. Long-term therapy for polymorphic mental disorders in liquidators of the consequences of the accident at the Chernobyl nuclear power plant

    Directory of Open Access Journals (Sweden)

    V. N. Krasnov

    2012-01-01

    Full Text Available The paper gives the results of a long-term comparative therapeutic study of a large cohort of more than 500 liquidators of the consequences of the accident at the Chernobyl nuclear power plant in 1986. The patients were followed up (and periodically treated at hospital 5 years or more, usually 10—15 years. The study confirmed mainly the cerebrovascular nature of disorders following the pattern seen in moderate psychoorganic syndrome. Therapy with cerebroprotective agents having vascular vegetotropic properties could yield certain therapeutic results and, to some extent, preserve social functioning capacity in these patients.

  15. Updated action plan for the implementation of measures as a consequence of the Fukushima reactor accident; Fortgeschriebener Aktionsplan zur Umsetzung von Massnahmen nach dem Reaktorunfall in Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-03-15

    The updated German action plan for the implementation of measures as a consequence of the Fukushima reactor accident covers the following issues: decision on the future utilization of nuclear energy in Germany; national frame for security checks, inspections and measures for nuclear power plants; international frame for inspections; action plan and WENRA reference level; action plan for the implementation of measures form robustness enhancement in German nuclear power plants (SNC topics 1-3), action plan for implementation of further measures (CNS topics 4-6).

  16. Accidents in nuclear ships

    Energy Technology Data Exchange (ETDEWEB)

    Oelgaard, P.L. [Risoe National Lab., Roskilde (Denmark)]|[Technical Univ. of Denmark, Lyngby (Denmark)

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

  17. 液氨储罐泄露后果模拟及应急处置%Consequences Simulation and Emergency Disposal of Liquid Ammonia Tank Leakage Accidents

    Institute of Scientific and Technical Information of China (English)

    孙高穹; 刘剑俊

    2016-01-01

    Taking a liquid ammonia tank of a enterprise cold storage as an example,This paper identified the risk of ammonia storage possible poisoning and explosion accident,used the TNT equivalent method and overpressure criterion to simulate the accident consequences of the steam explosion caused by the leakage of a single tank,predicted the toxic gas leakage diffusion effect based on multi-puff model,Finally put forward prevention and emergency response measures of the liquid ammonia tank leak accident.%文章以某企业冷库配备的液氨储罐为实例,对液氨储存过程可能发生的中毒和爆炸事故进行风险识别,运用TNT当量法和超压准则模拟单个储罐泄露后引发蒸汽云爆炸的事故后果,同时运用多烟团模式预测有毒气体泄露扩散影响范围,最后提出液氨储罐泄露事故防范和应急处置措施。

  18. A Hybrid Algorithm of Traffic Accident Data Mining on Cause Analysis

    Directory of Open Access Journals (Sweden)

    Jianfeng Xi

    2013-01-01

    Full Text Available Road traffic accident databases provide the basis for road traffic accident analysis, the data inside which usually has a radial, multidimensional, and multilayered structure. Traditional data mining algorithms such as association rules, when applied alone, often yield uncertain and unreliable results. An improved association rule algorithm based on Particle Swarm Optimization (PSO put forward by this paper can be used to analyze the correlation between accident attributes and causes. The new algorithm focuses on characteristics of the hyperstereo structure of road traffic accident data, and the association rules of accident causes can be calculated more accurately and in higher rates. A new concept of Association Entropy is also defined to help compare the importance between different accident attributes. T-test model and Delphi method were deployed to test and verify the accuracy of the improved algorithm, the result of which was a ten times faster speed for random traffic accident data sampling analyses on average. In the paper, the algorithms were tested on a sample database of more than twenty thousand items, each with 56 accident attributes. And the final result proves that the improved algorithm was accurate and stable.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-08-01

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

  20. [Paragliding accidents--a prospective analysis in Swiss mountain regions].

    Science.gov (United States)

    Lautenschlager, S; Karli, U; Matter, P

    1993-01-01

    During the period from 1.1 to 31.12.90, 86 injuries associated with paragliding were analysed in a prospective study in 12 different Swiss hospitals with reference to causes, patterns, and frequencies. Spine injuries (36%) and lesions of the lower extremities (35%) were diagnosed most frequently. Surprisingly no neurological complications occurred, which is possibly explained by the solitary axial trauma. In 15 cases very severe malleolar fractures required surgical intervention. One accident was fatal due to a lung rupture. 60% of all accidents happened during the landing phase, 26% at launching and 14% at flight. Half of the pilots were affected in their primary training course. Most accidents were due to an in-flight error of judgement, such as incorrect estimation of wind conditions and a choice of unfavourable landing sites. In contrast to early reports of hang-gliding injuries, only one accident was due to an equipment failure, namely a ruptured steering line. In more than a third of all accidents, the used paraglider was not in correct correlation with the pilot's weight and experience. Inspired by the desire for a long flight, gliders of too large surface-areas were often used, leading to a more unstable flight. To reduce the frequency of paragliding injuries, an accurate choice of equipment and increased attention to environmental factors is mandatory. Furthermore education-programs should focus more on intensifying the pilot's mental and practical skills.

  1. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    Science.gov (United States)

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

    1997-06-01

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

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

    Science.gov (United States)

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

    2004-01-01

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

  3. Severe Accident Analysis for Combustible Gas Risk Evaluation inside CFVS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, NaRae; Lee, JinYong; Bang, YoungSuk; Lee, DooYong [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Kim, HyeongTaek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The purpose of this study is to identify the composition of gases discharged into the containment filtered venting system by analyzing severe accidents. The accident scenarios which could be significant with respect to containment pressurization and hydrogen generation are derived and composition of containment atmosphere and possible discharged gas mixtures are estimated. In order to ensure the safety of the public and environment, the ventilation system should be designed properly by considering discharged gas flow rate, aerosol loads, radiation level, etc. One of considerations to be resolved is the risk due to combustible gas, especially hydrogen. Hydrogen can be generated largely by oxidation of cladding and decomposition of concrete. If the hydrogen concentration is high enough and other conditions like oxygen and steam concentration is met, the hydrogen can burn, deflagrate or detonate, which result in the damage the structural components. In particularly, after Fukushima accident, the hydrogen risk has been emphasized as an important contributor threatening the integrity of nuclear power plant during the severe accident. These results will be used to analyze the risk of hydrogen combustion inside the CFVS as boundary conditions. Severe accident simulation results are presented and discussed qualitatively with respect to hydrogen combustion. The hydrogen combustion risk inside of the CFVS has been examined qualitatively by investigating the discharge flow characteristics. Because the composition of the discharge flow to CFVS would be determined by the containment atmosphere, the severe accident progression and containment atmosphere composition have been investigated. Due to PAR operation, the hydrogen concentration in the containment would be decreased until the oxygen is depleted. After the oxygen is depleted, the hydrogen concentration would be increased. As a result, depending on the vent initiation timing (i.e. vent initiation pressure), the important

  4. A Study on the introduction of offsite consequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Kyu [Korea Environment Institute, Seoul (Korea)

    1998-12-01

    Considering accidents related to several chemical substances, it resulted in massive environmental disasters spreading to near region outside the accident site. Therefore, this study derived problems on the chemical substance accident management. It also presented a proposal for preventing or minimizing the effect on neighborhood and environment when an accident occurs. By reviewing several systems for reducing chemical substance accidents efficiently in the developed countries, a suitable system for Korea was suggested. 27 refs., 4 figs., 28 tabs.

  5. Analysis of traffic accidents on rural highways using Latent Class Clustering and Bayesian Networks.

    Science.gov (United States)

    de Oña, Juan; López, Griselda; Mujalli, Randa; Calvo, Francisco J

    2013-03-01

    One of the principal objectives of traffic accident analyses is to identify key factors that affect the severity of an accident. However, with the presence of heterogeneity in the raw data used, the analysis of traffic accidents becomes difficult. In this paper, Latent Class Cluster (LCC) is used as a preliminary tool for segmentation of 3229 accidents on rural highways in Granada (Spain) between 2005 and 2008. Next, Bayesian Networks (BNs) are used to identify the main factors involved in accident severity for both, the entire database (EDB) and the clusters previously obtained by LCC. The results of these cluster-based analyses are compared with the results of a full-data analysis. The results show that the combined use of both techniques is very interesting as it reveals further information that would not have been obtained without prior segmentation of the data. BN inference is used to obtain the variables that best identify accidents with killed or seriously injured. Accident type and sight distance have been identify in all the cases analysed; other variables such as time, occupant involved or age are identified in EDB and only in one cluster; whereas variables vehicles involved, number of injuries, atmospheric factors, pavement markings and pavement width are identified only in one cluster.

  6. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  7. Analysis of Criticality Accident Transients of Uranium Solution System

    Institute of Scientific and Technical Information of China (English)

    DUAN; Ming-hui; DU; Kai-wen; LIU; Zhen-hua

    2012-01-01

    <正>In the nuclear fuel cycle, fissile materials are often dissolved in water. Criticality accidents are likely to happen in the uranium solution system and release a large amount of energy and radioactive materials. Therefore, the criticality safety of uranium solution system is very important in the nuclear safety technology research.

  8. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  9. Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il, E-mail: sikim@kaeri.re.kr; Park, Jong Hwa; Ha, Kwang Soon; Cho, Song-Won; Song, JinHo

    2016-03-15

    Highlights: • Fukushima unit 2 accident was analyzed using MELCOR 1.8.6. • RCIC operating conditions were assumed and best case was selected. • Effect of RCIC operating condition on accident scenario was found. - Abstract: A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-30

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

  11. What are the consequences of the reactor accident in Fukushima for the evaluation of nuclear risk?; Welche Folgen hat der Kernkraftwerksunfall in Fukushima fuer die Bewertung von Kernenergierisiken?

    Energy Technology Data Exchange (ETDEWEB)

    Renn, Ortwin [Univ. Stuttgart (Germany). Inst. Sozialwissenshaften V; Gallego Carrera, Diana [Univ. Stuttgart (Germany). ZIRIUS Zentrum fuer Interdiszipliaere Risiko- und Innovationsforschung

    2015-06-01

    There are historical breaks in the relation of risk analysis, risk perception and regulation policy. The year 2011 with the reactor accident in the NPP Fukushima was such a break, especially in Germany. The nuclear phase-out was reduced to ten years the energy policy turnaround received a broad societal agreement. Nuclear facilities loose public acceptance, the risk perception has changed. The Japanese evaluation results on faulty and nontransparent behavior and the lack of governance of responsible persons and authorities including a poor accident management have further decreased the public confidence. A new concept of safety culture for all nuclear facilities including the radioactive waste management is required, the communication processes between plant operator, authorities, science and the public have to be intensified.

  12. A GIS-based prediction and assessment system of off-site accident consequence for Guangdong nuclear power plant.

    Science.gov (United States)

    Wang, X Y; Qu, J Y; Shi, Z Q; Ling, Y S

    2003-01-01

    GNARD (Guangdong Nuclear Accident Real-time Decision support system) is a decision support system for off-site emergency management in the event of an accidental release from the nuclear power plants located in Guangdong province, China. The system is capable of calculating wind field, concentrations of radionuclide in environmental media and radiation doses. It can also estimate the size of the area where protective actions should be taken and provide other information about population distribution and emergency facilities available in the area. Furthermore, the system can simulate and evaluate the effectiveness of countermeasures assumed and calculate averted doses by protective actions. All of the results can be shown and analysed on the platform of a geographical information system (GIS).

  13. A common sense approach to consequence analysis at a large DOE site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    O`Kula, K.R.; McKinley, M.S.; East, J.M.

    1992-12-31

    The primary objective of the Probabilistic Safety Assessment (PSA) at the U. S. Department of Energy (DOE) Savannah River Site (SRS) is to quantify health and economic risks posed by K Reactor operation to the nearby offsite and onsite areas from highly unlikely severe accidents. The overall risk analyses have also been instrumental as defensible bases for analyzing existing safety margins of the restart configuration; determining component, human action, and engineering system vulnerabilities; comparing measures of risk to DOE and commercial guidelines; and prioritizing risk-significant improvements. The key final phase of these probabilistic risk calculations, a third level of analysis or Level 3 PSA, requires the determination of the conditional consequences to onsite workers and the DOE reservation facilities, given low-probability, postulated fuel-melting accidents with accompanying atmospheric releases have occurred. A modified version of the commercial reactor-based MACCS 1.5 code, MACCS/ON, is used in the context of the SRS PSA to perform the consequence determinations. The updated code is applicable to other large DOE sites for risk analyses of facility operations, and is compatible with proposed modifications planned by code developers, Sandia National Laboratories.

  14. A common sense approach to consequence analysis at a large DOE site

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; McKinley, M.S.; East, J.M.

    1992-01-01

    The primary objective of the Probabilistic Safety Assessment (PSA) at the U. S. Department of Energy (DOE) Savannah River Site (SRS) is to quantify health and economic risks posed by K Reactor operation to the nearby offsite and onsite areas from highly unlikely severe accidents. The overall risk analyses have also been instrumental as defensible bases for analyzing existing safety margins of the restart configuration; determining component, human action, and engineering system vulnerabilities; comparing measures of risk to DOE and commercial guidelines; and prioritizing risk-significant improvements. The key final phase of these probabilistic risk calculations, a third level of analysis or Level 3 PSA, requires the determination of the conditional consequences to onsite workers and the DOE reservation facilities, given low-probability, postulated fuel-melting accidents with accompanying atmospheric releases have occurred. A modified version of the commercial reactor-based MACCS 1.5 code, MACCS/ON, is used in the context of the SRS PSA to perform the consequence determinations. The updated code is applicable to other large DOE sites for risk analyses of facility operations, and is compatible with proposed modifications planned by code developers, Sandia National Laboratories.

  15. Preliminary accident analysis of Flexblue® underwater reactor

    Directory of Open Access Journals (Sweden)

    Haratyk Geoffrey

    2015-01-01

    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  16. Evaluation of sanitary consequences of Chernobyl accident in France: epidemiological monitoring device, state of knowledge, evaluation of risks and perspectives; Evaluation des consequences sanitaires de l'accident de Tchernobyl en France: dispositif de surveillance epidemiologique, etat des connaissances, evaluation des risques et perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Verger, P.; Champion, D.; Gourmelon, P.; Hubert, Ph.; Joly, J.; Renaud, Ph.; Tirmarche, M.; Vidal, M. [CEA/Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, IPSN, 92 (France); Cherie-Challine, L.; Boutou, O.; Isnard, H.; Jouan, M.; Pirard, Ph. [Institut National de Veille Sanitaire, 94 - Saint-Maurice (France)

    2000-12-01

    The objectives of this document are firstly, to present the situation of knowledge both on the sanitary consequences of the Chernobyl accident and on the risk factors of thyroid cancers, these ones constituting one of the most principal consequences observed in Belarus, in Ukraine and Russia; secondly, the give the principal system contributing to the epidemiological surveillance of effects coming from a exposure to ionizing radiations, in France and to give the knowledge on incidence and mortality of thyroid cancer in France; thirdly, to discuss the pertinence and the feasibility of epidemiological approaches that could be considered to answer questions that the public and authorities ask relatively to the sanitary consequences of Chernobyl accident in France; fourthly to male a calculation of thyroid cancer risk in relation with Chernobyl fallout in France from works and studies made from 1986 on the consequences of this disaster in terms of radioecology and dosimetry at the national level. Besides, the improvement of thyroid cancer surveillance is also tackled. (N.C.)

  17. RASCAL 及其在核事故后果评价中的应用%RASCAL and Its Application in Nuclear Accident Consequences Assessment

    Institute of Scientific and Technical Information of China (English)

    王韶伟; 侯杰; 陈海英; 曹亚丽; 乔清党; 李冰

    2014-01-01

    The development history, main function and basic principle used for emergency response of RAS-CAL, which is used for analyzing nuclear and radiate accident by American Nuclear Regulatory Commission, are presented in the study.The main modules/models are analyzed selectively, including source term to dose, field measurement to dose, meteorological data processor, source term calculation, transport, diffusion, and dose calculations.Then, RASCAL is applied to assess the radiological consequence of a nuclear power plant ac-cident emergency exercise.The assessment conclusion is displayed through Google Earth as 3D style.%介绍了美国核管会用于核与辐射事故后果分析的辐射评价系统( RASCAL)的主要功能和特性,重点分析了RASCAL的源项计算剂量模块、场外监测数据计算剂量模块、气象数据处理模块,以及源项计算模式、大气输运扩散模式和剂量计算模式。最后,将RASCAL应用于我国某核电厂事故应急演习中,评价分析事故情景下的放射性影响,并将其结果通过Google Earth进行三维展示。

  18. SCIENTIFIC SUPPORT OF THE MEDICAL SECTION OF THE STATE PROGRAM OF THE BELARUS REPUBLIC FOR THE OVERCOMING OF THE CHERNOBYL ACCIDENT CONSEQUENCES

    Directory of Open Access Journals (Sweden)

    A. V. Rozhko

    2012-01-01

    Full Text Available A twenty-five year health follow-up of the affected population has shown that a properly structured State strategy on overcoming the consequences of disaster allow to maintain stable levels of morbidity and mortality. An important achievement in the system of medical help to the affected population is the organization of dynamic follow-up, as well as creating State Register of people exposed to radiation as a result of the Chernobyl accident as a tool for solving scientific and practical problems. The results of scientific researches obtained in the SO “The Republican Research Centre for Radiation Medicine and Human Ecology” were the basis for one of the Council of Ministers Decree and two Decrees of the Ministry of Health. Significant changes have been made in the order of assigning the causation connection of disease (disability and the accident at the Chernobyl nuclear power plant and objective criteria for the formation of high radiation risk groups.In a whole, the rate of oncological morbidity in the affected population remains at the average republican level, but for certain categories of the affected population, referred to groups of enhanced radiation risk, there has been detected the presence of excess morbidity of some forms of malignant neoplasms.

  19. Consequences of Windscale accident (October 1957) and study of the validity of the Sutton's mathematical model of atmospheric diffusion (1960); Etude des consequences de l'accident de Windscale (Octobre 1957) et de la validite du modele mathematique de diffusion atmospherique de Sutton (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Doury, A. [Commissariat a l' Energie Atomique (S.C.R.G.R.) Saclay (France).Centre d' Etudes Nucleaires; Martin, J.J. [Electricite de France (EDF)(S.L.P.R.), 37 - Chinon (France)

    1960-07-01

    The reactor accident that happens at the number 1 pile of Windscale in 1957 was followed by a discharge of radioactive products into the atmosphere from the 1.X.1957 at 4.30 PM to the 12.X.1957 at 3.10 PM. On october the 11{sup th} it was possible to say that there was no more risk either of external irradiation or inhalation. But in adopting a M.A.C. of 0,1 {mu}curie of iodine 131 per litre of milk, the Authority had to control the milk delivery till november 23{sup rd} on a 500 km{sup 2} area. On the other hand, this exceptional accident permit to verify that Sutton's atmospheric diffusion model could give an easy means to foresee, with a sufficient approximation, the consequences of a dispersion of radioactive products into the atmosphere. (author) [French] L'accident survenu a la pile numero 1 de Windscale en 1957 a entraine l'emission de matieres radioactives dans l'atmosphere du 10 octobre a 16h30 au 12 octobre a 15h10. Le 11 octobre, on pouvait dire qu'il n'y avait plus de risque d'irradiation externe ni de danger par inhalation. Mais en adoptant une C.M.A. de 0,1 {mu}curie d'iode 131 par litre de lait, les autorites ont du reglementer la consommation du lait jusqu'au 23 novembre sur une etendue d'environ 500 km{sup 2}. D'autre part, cet accident exceptionnel a permis de verifier que le modele de diffusion atmospherique de Sutton pouvait fournir un moyen commode de prevoir avec une approximation suffisante les consequences d'une dispersion de produits radioactifs dans l'atmosphere. (auteur)

  20. Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

  1. Analysis of Core Degradation in Fukushima Unit 1 Accident with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Kim, Tae Woon; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    In this study, an accident analysis of Fukushima Daiichi Unit 1 was performed using MELCOR 1.8.6. The behavior of the initial stage of the accident was focused during 30 hours after the reactor scram, because it was predicted that the vessel failure (severe accident) occurred before 20 hours. A hydrogen explosion also occurred at about 24 hours after the accident, and thus the phenomenon of core degradation before 30 hours was highlighted. Moreover, the effect of the amount of fresh water injection on the core degradation was performed by changing the amount of injection water. It was expected that a large portion of the injection water could not reach the core because of leakage. Thus, core damage was observed according to the amount of water that reached the core. The plant geometries and operating conditions were obtained from TEPCO (Tokyo Electric Power Company) through the OECD/NEA BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) Project. An analysis of the Fukushima accident was also performed by Sandia National Laboratories, but several conditions were revised and added in this study. First, the flow rate of the steam into turbine and water into downcomer were considered at the initial stage of accident. The water level followed well the measured data by adding this mechanism. Second, SRV stuck open was included in this calculation. SRV stuck open can occur due to high temperature and frequent operation, and was modeled in calculation. The timing of SRV stuck open was closed to the timing of MSL failure, and thus the depressurization of RPV could have originated from both of MSL failure and SRV stuck open. The effect of injection water was observed, and it was found that the proper water injection can prevent a severe accident at the initial stage of the accident. In conclusion, an analysis of the severe accident occurring in Fukushima Unit 1 was conducted by using MELCOR. The analysis results were consistent with the

  2. An aftermath analysis of the 2014 coal mine accident in Soma, Turkey: Use of risk performance indicators based on historical experience.

    Science.gov (United States)

    Spada, Matteo; Burgherr, Peter

    2016-02-01

    On the 13th of May 2014 a fire related incident in the Soma coal mine in Turkey caused 301 fatalities and more than 80 injuries. This has been the largest coal mine accident in Turkey, and in the OECD country group, so far. This study investigated if such a disastrous event should be expected, in a statistical sense, based on historical observations. For this purpose, PSI's ENSAD database is used to extract accident data for the period 1970-2014. Four different cases are analyzed, i.e., OECD, OECD w/o Turkey, Turkey and USA. Analysis of temporal trends for annual numbers of accidents and fatalities indicated a non-significant decreasing tendency for OECD and OECD w/o Turkey and a significant one for USA, whereas for Turkey both measures showed an increase over time. The expectation analysis revealed clearly that an event with the consequences of the Soma accident is rather unlikely for OECD, OECD w/o Turkey and USA. In contrast, such a severe accident has a substantially higher expectation for Turkey, i.e. it cannot be considered an extremely rare event, based on historical experience. This indicates a need for improved safety measures and stricter regulations in the Turkish coal mining sector in order to get closer to the rest of OECD.

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

    Directory of Open Access Journals (Sweden)

    Mehran Amiri

    2015-04-01

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

  4. Analysis of swimming pool accidents resulting in spinal cord injury.

    Science.gov (United States)

    Green, B A; Gabrielsen, M A; Hall, W J; O'Heir, J

    1980-04-01

    This paper is a summary of a study of 72 cases of swimming-pool accidents resulting in serious injuries with the potential of permanent disability. Sixty-four of the 72 cases resulted in spinal cord injuries, 57 of which involved quadriplegic lesions. The authors observed that the majority of these injuries resulted from a lack of good judgement and common sense rather than from intoxication or pool structural deficiencies. Also of note was the lack of appropriate first-aid and extrication rendered, as well as the absence of uniform treatment and care received by the majority of the patients.

  5. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Ha, Kwang Soon; Kim, Hwan-Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

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

    Science.gov (United States)

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

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

  7. The accident in Fukushima. Preliminary report on the accident progress in the nuclear power plants as a consequence of the earth quake on 11th March 2011; Der Unfall in Fukushima. Zwischenbericht zu den Ablaeufen in den Kernkraftwerken nach dem Erdbeben vom 11. Maerz 2011

    Energy Technology Data Exchange (ETDEWEB)

    Borghoff, Stefan; Brueck, Benjamin; Kilian-Huelsmeyer, Yvonne; Maqua, Michael; Mildenberger, Oliver; Quester, Claudia; Stahl, Thorsten; Thuma, Gernot; Wetzel, Norbert; Wild, Volker

    2011-08-15

    The preliminary report on the accident progress in the nuclear power plants as a consequence of the earth quake on 11th March 2011 describes the chronologic sequence of the accident in the different units of the power plant. The measures for mitigation of the accident impact at the site of Fukushima Daiichi and Fukushima Daini included the efforts to reach and maintain stable plant conditions. The issue radiological situation includes an estimation of the air-borne radionuclide release, the contamination of the environment and the sea water, measures for protection of the public. The lessons learned following the NISA and IAEA fact finding missions and the open questions are summarized.

  8. ASTEC adaptation for PHWR limited core damage accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Chatterjee, B.; Lele, H.G. [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Guillard, G.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, 13115 Saint-Paul-lez-Durance (France)

    2014-06-01

    Under limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1.

  9. [Model of Analysis and Prevention of Accidents - MAPA: tool for operational health surveillance].

    Science.gov (United States)

    de Almeida, Ildeberto Muniz; Vilela, Rodolfo Andrade de Gouveia; da Silva, Alessandro José Nunes; Beltran, Sandra Lorena

    2014-12-01

    The analysis of work-related accidents is important for accident surveillance and prevention. Current methods of analysis seek to overcome reductionist views that see these occurrences as simple events explained by operator error. The objective of this paper is to analyze the Model of Analysis and Prevention of Accidents (MAPA) and its use in monitoring interventions, duly highlighting aspects experienced in the use of the tool. The descriptive analytical method was used, introducing the steps of the model. To illustrate contributions and or difficulties, cases where the tool was used in the context of service were selected. MAPA integrates theoretical approaches that have already been tried in studies of accidents by providing useful conceptual support from the data collection stage until conclusion and intervention stages. Besides revealing weaknesses of the traditional approach, it helps identify organizational determinants, such as management failings, system design and safety management involved in the accident. The main challenges lie in the grasp of concepts by users, in exploring organizational aspects upstream in the chain of decisions or at higher levels of the hierarchy, as well as the intervention to change the determinants of these events.

  10. An Analysis of Human Causal Factors in Unmanned Aerial Vehicle (UAV) Accidents

    Science.gov (United States)

    2014-12-01

    NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA MBA PROFESSIONAL REPORT AN ANALYSIS OF HUMAN CAUSAL FACTORS IN UNMANNED AERIAL...December 2014 3. REPORT TYPE AND DATES COVERED MBA Professional Report 4. TITLE AND SUBTITLE AN ANALYSIS OF HUMAN CAUSAL FACTORS IN UNMANNED...287 causal factors attributed to 68 accidents, 65 percent of the factors were associated with humans. Moreover, this study also discloses that the

  11. 90Sr and 89Sr in seawater off Japan as a consequence of the Fukushima Dai-ichi nuclear accident

    Directory of Open Access Journals (Sweden)

    K. O. Buesseler

    2013-02-01

    Full Text Available The impact of the earthquake and tsunami in the east coast of Japan in 11 March 2011 caused a loss of power at the Fukushima Dai-ichi Nuclear Power Plant (NPP that resulted in one of the most important releases of artificial radioactivity to the environment. Although several works were devoted to evaluate the atmospheric dispersion of radionuclides, the impact of the discharges to the ocean has been less investigated. Here we evaluate the distribution of Fukushima-derived 90Sr and 89Sr throughout waters 30–600 km offshore in June 2011. Concentrations of 90Sr and 89Sr in both surface waters and shallow profiles ranged from 0.8 ± 0.2 to 85 ± 3 Bq m−3 and from 19 ± 6 to 265 ± 74 Bq m−3, respectively. Because of its short half-life, all measured 89Sr was due to the accident, while the 90Sr concentrations can be compared to the background levels in the Pacific Ocean of about 1.2 Bq m−3. Fukushima-derived radiostrontium was mainly detected north of Kuroshio Current, as this was acting as a southern boundary for transport. The highest activities were associated with near-shore eddies, and larger inventories were found in the closest stations to Fukushima NPP. The data evidences a major influence of direct liquid discharges of radiostrontium compared to the atmospheric deposition. Existing 137Cs data reported from the same samples allowed us establishing a 90Sr/137Cs ratio of 0.0256 ± 0.0006 in seawater off Fukushima, being significantly different than that of the global atmospheric fallout (i.e. 0.63 and may be used in future studies to track waters coming from the east coast of Japan. Liquid discharges of 90Sr to the ocean were estimated, resulting in an inventory of 53 ± 1 TBq of 90Sr in the inshore study area in June 2011 and total releases of 90Sr ranging from 90 to 900 TBq, depending upon the reported estimates of 137Cs releases that are considered.

  12. 90Sr and 89Sr in seawater off Japan as a consequence of the Fukushima Dai-ichi nuclear accident

    Directory of Open Access Journals (Sweden)

    N. Casacuberta

    2013-06-01

    Full Text Available The impact of the earthquake and tsunami on the east coast of Japan on 11 March 2011 caused a loss of power at the Fukushima Dai-ichi nuclear power plant (NPP that resulted in one of the most important releases of artificial radioactivity into the environment. Although several works were devoted to evaluating the atmospheric dispersion of radionuclides, the impact of the discharges to the ocean has been less investigated. Here we evaluate the distribution of Fukushima-derived 90Sr (n = 57 and 89Sr (n = 19 throughout waters 30–600 km offshore in June 2011. Concentrations of 90Sr and 89Sr in both surface waters and shallow profiles ranged from 0.8 ± 0.2 to 85 ± 3 Bq m−3 and from 19 ± 6 to 265 ± 74 Bq m−3, respectively. Because of its short half-life, all measured 89Sr was due to the accident, while the 90Sr concentrations can be compared to the background levels in the Pacific Ocean of about 1.2 Bq m−3. Fukushima-derived radiostrontium was mainly detected north of Kuroshio Current, as this was acting as a southern boundary for transport. The highest activities were associated with near-shore eddies, and larger inventories were found in the closest stations to Fukushima NPP. The data evidence a major influence of direct liquid discharges of radiostrontium compared to the atmospheric deposition. Existing 137Cs data reported from the same samples allowed us to establish a 90Sr / 137Cs ratio of 0.0256 ± 0.0006 in seawater off Fukushima, being significantly different than that of the global atmospheric fallout (i.e., 0.63 and may be used in future studies to track waters coming from the east coast of Japan. Liquid discharges of 90Sr to the ocean were estimated, resulting in an inventory of 53 ± 1 TBq of 90Sr in the inshore study area in June 2011 and total releases of 90Sr ranging from 90 to 900 TBq, depending upon the reported estimates of 137Cs releases that are considered.

  13. 90Sr and 89Sr in seawater off Japan as a consequence of the Fukushima Dai-ichi nuclear accident

    Science.gov (United States)

    Casacuberta, N.; Masqué, P.; Garcia-Orellana, J.; Garcia-Tenorio, R.; Buesseler, K. O.

    2013-06-01

    The impact of the earthquake and tsunami on the east coast of Japan on 11 March 2011 caused a loss of power at the Fukushima Dai-ichi nuclear power plant (NPP) that resulted in one of the most important releases of artificial radioactivity into the environment. Although several works were devoted to evaluating the atmospheric dispersion of radionuclides, the impact of the discharges to the ocean has been less investigated. Here we evaluate the distribution of Fukushima-derived 90Sr (n = 57) and 89Sr (n = 19) throughout waters 30-600 km offshore in June 2011. Concentrations of 90Sr and 89Sr in both surface waters and shallow profiles ranged from 0.8 ± 0.2 to 85 ± 3 Bq m-3 and from 19 ± 6 to 265 ± 74 Bq m-3, respectively. Because of its short half-life, all measured 89Sr was due to the accident, while the 90Sr concentrations can be compared to the background levels in the Pacific Ocean of about 1.2 Bq m-3. Fukushima-derived radiostrontium was mainly detected north of Kuroshio Current, as this was acting as a southern boundary for transport. The highest activities were associated with near-shore eddies, and larger inventories were found in the closest stations to Fukushima NPP. The data evidence a major influence of direct liquid discharges of radiostrontium compared to the atmospheric deposition. Existing 137Cs data reported from the same samples allowed us to establish a 90Sr / 137Cs ratio of 0.0256 ± 0.0006 in seawater off Fukushima, being significantly different than that of the global atmospheric fallout (i.e., 0.63) and may be used in future studies to track waters coming from the east coast of Japan. Liquid discharges of 90Sr to the ocean were estimated, resulting in an inventory of 53 ± 1 TBq of 90Sr in the inshore study area in June 2011 and total releases of 90Sr ranging from 90 to 900 TBq, depending upon the reported estimates of 137Cs releases that are considered.

  14. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  15. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  16. Narrative Text Analysis of Accident Reports with Tractors, Self-Propelled Harvesting Machinery and Materials Handling Machinery in Austrian Agriculture from 2008 to 2010 – A Comparison

    OpenAIRE

    Hannes Mayrhofer; Elisabeth Quendler; Josef Boxberger

    2014-01-01

    The aim of this study was the identification of accident scenarios and causes by analysing existing accident reports of recognized agricultural occupational accidents with tractors, self-propelled harvesting machinery and materials handling machinery from 2008 to 2010. As a result of a literature-based evaluation of past accident analyses, the narrative text analysis was chosen as an appropriate method. A narrative analysis of the text fields of accident reports that farmers used to report ac...

  17. The Research Summary of UF6 Leakage Accident Consequence Assessment%UF6泄漏事故后果评价研究进展

    Institute of Scientific and Technical Information of China (English)

    陈海龙

    2013-01-01

    简要介绍了核燃料循环过程中UF6泄漏事故的几类事故情形,以及UF6泄漏后的大气扩散过程。目前,用于UF6泄漏事故后果评价的主要模型是HGSYSTEM/UF6模型和RASCAL模型:一般情况下,两种模型可溶性铀的平均浓度的预测值与实际测量值相比为小于2的数;在D类稳定下RASCAL预测的结果处于高斯模型和HGSYSTEM/UF6之间;而在F类稳定度下,1 km内基本上是RASCAL计算结果最低,1 km外3个模型预测结果无规律性。%This paper gathered and analyzed the scenarios for UF6 release accident ,and the atmosphere diffu-sion of UF6 and the air concentration of UF6 ,HF ,UO2 F2 .The result showed that two models are used to sim-ulate UF6 accidental release now , HGSYSTEM/UF6 model and RASCAL model . The former is a dense gas model ,and the latter is an accident consequence assessment model .The scholar of Portugal compared the two different dispersion models .The results showed that the calculated concentration by RASCAL model is between the HGSYSTEM/UF6 model and Gaussian plume model under the D stability ,but the tendency is complicated under the F stability .

  18. Improvement of radiological consequence estimation methodologies for NPP accidents in the ARGOS and RODOS decision support systems through consideration of contaminant physico-chemical forms

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, K.G.; Roos, P. [Technical University of Denmark - DTU (Denmark); Lind, O.C.; Salbu, B. [Norwegian University of Life Sciences/CERAD - NMBU (Norway); Bujan, A.; Duranova, T. [VUJE, Inc. (Slovakia); Ikonomopoulos, A.; Andronopoulos, S. [National Centre for Scientific Research ' Demokritos' (Greece)

    2014-07-01

    The European standard computerized decision support systems RODOS and ARGOS, which are integrated in the operational nuclear emergency preparedness in practically all European countries, as well as in a range of non-European countries, are highly valuable tools for radiological consequence estimation, e.g., in connection with planning and exercising as well as in justification and optimization of intervention strategies. Differences between the Chernobyl and Fukushima accident atmospheric release source terms have demonstrated that differences in release conditions and processes may lead to very different degrees of volatilization of some radionuclides. Also the physico-chemical properties of radionuclides released can depend strongly on the release process. An example from the Chernobyl accident of the significance of this is that strontium particles released in the fire were oxidized and thus generally physico-chemically different from those released during the preceding explosion. This is reflected in the very different environmental mobility of the two groups of particles. The initial elemental matrix characteristics of the contaminants, as well as environmental parameters like pH, determine for instance the particle dissolution time functions, and thus the environmental mobility and potential for uptake in living organisms. As ICRP recommends optimization of intervention according to residual dose, it is crucial to estimate long term dose contributions adequately. In the EURATOM FP7 project PREPARE, an effort is made to integrate physico-chemical forms of contaminants in scenario-specific source term determination, thereby enabling consideration of influences on atmospheric dispersion/deposition, post-deposition migration, and effectiveness of countermeasure implementation. The first step in this context was to investigate, based on available experience, the important physico-chemical properties of radio-contaminants that might potentially be released to the

  19. Economic consequences of aviation system disruptions: A reduced-form computable general equilibrium analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhenhua; Rose, Adam Z.; Prager, Fynnwin; Chatterjee, Samrat

    2017-01-01

    The state of the art approach to economic consequence analysis (ECA) is computable general equilibrium (CGE) modeling. However, such models contain thousands of equations and cannot readily be incorporated into computerized systems used by policy analysts to yield estimates of economic impacts of various types of transportation system failures due to natural hazards, human related attacks or technological accidents. This paper presents a reduced-form approach to simplify the analytical content of CGE models to make them more transparent and enhance their utilization potential. The reduced-form CGE analysis is conducted by first running simulations one hundred times, varying key parameters, such as magnitude of the initial shock, duration, location, remediation, and resilience, according to a Latin Hypercube sampling procedure. Statistical analysis is then applied to the “synthetic data” results in the form of both ordinary least squares and quantile regression. The analysis yields linear equations that are incorporated into a computerized system and utilized along with Monte Carlo simulation methods for propagating uncertainties in economic consequences. Although our demonstration and discussion focuses on aviation system disruptions caused by terrorist attacks, the approach can be applied to a broad range of threat scenarios.

  20. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  1. Analysis on the nitrogen drilling accident of Well Qionglai 1 (I: Major inducement events of the accident

    Directory of Open Access Journals (Sweden)

    Yingfeng Meng

    2015-12-01

    Full Text Available Nitrogen drilling in poor tight gas sandstone should be safe because of very low gas production. But a serious accident of fire blowout occurred during nitrogen drilling of Well Qionglai 1. This is the first nitrogen drilling accident in China, which was beyond people's knowledge about the safety of nitrogen drilling and brought negative effects on the development of gas drilling technology still in start-up phase and resulted in dramatic reduction in application of gas drilling. In order to form a correct understanding, the accident was systematically analyzed, the major events resulting in this accident were inferred. It is discovered for the first time that violent ejection of rock clasts and natural gas occurred due to the sudden burst of downhole rock when the fractured tight gas zone was penetrated during nitrogen drilling, which has been named as “rock burst and blowout by gas bomb”, short for “rock burst”. Then all the induced events related to the rock burst are as following: upthrust force on drilling string from rock burst, bridging-off formed and destructed repeatedly at bit and centralizer, and so on. However, the most direct important event of the accident turns out to be the blockage in the blooie pipe from rock burst clasts and the resulted high pressure at the wellhead. The high pressure at the wellhead causes the blooie pipe to crack and trigged blowout and deflagration of natural gas, which is the direct presentation of the accident.

  2. The Cost of Company Occupational Accidents: An Activity Based Analysis using the SACA Method

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.; Impgaard, Martin

    The Systematic Accident Cost Analysis (SACA) project is a research project carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. It empirically tested a method - the SACA method...... - for evaluating the visible and hidden costs of corporate occupational accidents. It also focused on whether the registration, processing and reporting of these costs could be integrated in the corporate accounting information system. The project was based on case studies in 9 Danish companies within 3 different...

  3. Study on coal mines accidents based on the grey relational analysis

    Institute of Scientific and Technical Information of China (English)

    WANG Shuai; ZHANG Jin-long

    2008-01-01

    The subject investigated the system of people-machine-environment in coal mines. The coal mines working process was researched and the theory of grey system was applied to analyze coal mines safety accidents and those relevant factors. This re-search reveals that this analysis method is easy and highly available and the result is of great credibility, which can not only provide theoretical supports to the quantitative study of coal mines safety accident, but offer basis for coal mines companies' safety management.

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

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-15

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

  5. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    CERN Document Server

    Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

    2013-01-01

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

  6. Safety Analysis in Large Volume Vacuum Systems Like Tokamak: Experiments and Numerical Simulation to Analyze Vacuum Ruptures Consequences

    Directory of Open Access Journals (Sweden)

    A. Malizia

    2014-01-01

    Full Text Available The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of hazardous internal material through the braches. It is clear that the influence of flow fields, consequence of accidents like LOVA, on dust resuspension is a key safety issue. In order to develop this analysis an experimental facility is been developed: STARDUST. This last facility has been used to improve the knowledge about LOVA to replicate a condition more similar to appropriate operative condition like to kamaks. By the experimental data the boundary conditions have been extrapolated to give the proper input for the 2D thermofluid-dynamics numerical simulations, developed by the commercial CFD numerical code. The benchmark of numerical simulation results with the experimental ones has been used to validate and tune the 2D thermofluid-dynamics numerical model that has been developed by the authors to replicate the LOVA conditions inside STARDUST. In present work, the facility, materials, numerical model, and relevant results will be presented.

  7. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H{sub 2}/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author).

  8. Analysis of dental materials as an aid to identification in aircraft accidents

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.S.; Cruickshanks-Boyd, D.W.

    1982-04-01

    The failure to achieve positive identification of aircrew following an aircraft accident need not prevent a full autopsy and toxicological examination to ascertain possible medical factors involved in the accident. Energy-dispersive electron microprobe analysis provides morphological, qualitative, and accurate quantitative analysis of the composition of dental amalgam. Wet chemical analysis can be used to determine the elemental composition of crowns, bridges and partial dentures. Unfilled resin can be analyzed by infrared spectroscopy. Detailed analysis of filled composite restorative resins has not yet been achieved in the as-set condition to permit discrimination between manufacturers' products. Future work will involve filler studies and pyrolysis of the composite resins by thermogravimetric analysis to determine percentage weight loss when the sample examined is subjected to a controlled heating regime. With these available techniques, corroborative evidence achieved from the scientific study of materials can augment standard forensic dental results to obtain a positive identification.

  9. Analysis of Adolescent Awareness of Radiation: Marking the First Anniversary of the Fukushima Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bang Ju [Korean Science Reporters Association, Seoul (Korea, Republic of)

    2012-06-15

    Marking the first anniversary of the Fukushima nuclear accident, which took place on March 11th, 2011, the level of adolescent awareness and understanding of radiation was surveyed, and the results were then compared with those for adults with the same questionnaires conducted at similar times. A qualitative survey and frequency analysis were made for the design of the study methodology. Those surveyed were limited to 3rd grade middle school students, 15 years of age, who are the future generation. The questionnaire, which is a survey tool, was directly distributed to the students and 2,217 answers were analysed. The questionnaires were composed of 40 questions, and it was found that Cronbach's coefficient was high with 'self awareness of radiation' at 0.494, 'risk of radiation' at 0.843, 'benefit of radiation' at 0.748, 'radiological safety control' at 0.692, 'information sources of radiation' at 0.819, and 'impacts of Fukushima accident'. The results of the survey analysis showed that the students' knowledge of radiation was not very high with 67.4 points (69.5 points for adults) calculated on a maximum scale of 100 points (converted points). The impacts of the Fukushima nuclear accident were found to be less significant to adolescents than adults, and the rate of answer of 'so' or ' very so' in the following questions demonstrates this well. It was also shown that the impacts of the Fukushima accident to adolescents were comparatively low with 27.0% (38.9% for adults) on the question of 'attitude changed against nuclear power due to the Fukushima accident,' 65.7%(86.6% for adults) on the question of 'the damages from the Fukushima accident was immeasurably huge,' and 65.0% (86.3% for adults) on 'the Fukushima accident contributed to raising awareness on the safety of nuclear power plants'. The adolescents had a high rate of &apos

  10. Resolve! Version 2.5: Flammable Gas Accident Analysis Tool Acceptance Test Plan and Test Results

    Energy Technology Data Exchange (ETDEWEB)

    LAVENDER, J.C.

    2000-10-17

    RESOLVE! Version 2 .5 is designed to quantify the risk and uncertainty of combustion accidents in double-shell tanks (DSTs) and single-shell tanks (SSTs). The purpose of the acceptance testing is to ensure that all of the options and features of the computer code run; to verify that the calculated results are consistent with each other; and to evaluate the effects of the changes to the parameter values on the frequency and consequence trends associated with flammable gas deflagrations or detonations.

  11. Formal Analysis of an Airplane Accident in N{Σ}-Labeled Calculus

    Science.gov (United States)

    Mizutani, Tetsuya; Igarashi, Shigeru; Ikeda, Yasuwo; Shio, Masayuki

    N{Σ}-labeled calculus is a formal system for representation, verification and analysis of time-concerned recognition, knowledge, belief and decision of humans or computer programs together with related external physical or logical phenomena.In this paper, a formal verification and analysis of the JAL near miss accident is presented as an example of cooperating systems controlling continuously changing objects including human factor with misunderstanding or incorrect recognition.

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

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  13. Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

    2016-09-15

    At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

  14. DATA ANALYSIS OF TRAFFIC ACCIDENTS AND THEIR CAUSES IN GOMEL FOR 2013 AND 2014

    Directory of Open Access Journals (Sweden)

    S. A. Azemsha

    2015-01-01

    Full Text Available Measures undertaken for improvement of road traffic safety presuppose clampdown on violation of road traffic regulations but no attention is paid to the fact that there is no possibility to ensure road traffic safety due to inconformity of roads to the modern safety requirements. Therefore reduction in accident rate is connected with some changes in approaches to designing, construction and maintenance of roads. Nowadays when the number of automobiles is extremely large and their number is increasing with every passing year driver’s professionalism becomes the most significant factor. In these circumstances the professionalism is demonstrated not so much while driving in bad road conditions as it was previously but the professionalism is more revealed in the case when it is necessary to drive in conditions of large workloads and high manoeuvring rate when a special important role is given to the capability to forecast a situation, in other words the capability to read the road. Moral climate on the road is no less important as well and it practically fully depends on a driver.The paper contains an analysis of the Gomel traffic police data on quantitative distribution of road traffic accidents and their victims according to the time of day and month, lighting conditions, weather conditions, age and sex factors, types of violations. Situations of traffic behavior, motivations of drivers and road users, drivers training have been shown in terms of impact on the road traffic accident risk. The paper considers a human factor which rather often causes an accident. An evaluation has been given to such factor as road speed of transport facilities which exerts a significant influence on an accident risk and its severity.

  15. ANALYSIS OF MEDIA PUBLICATIONS ON THE FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT

    Directory of Open Access Journals (Sweden)

    I. A. Zykova

    2011-01-01

    Full Text Available The analysis of informing the public about radiation and radiation risks is made on the basiс of publications on the Fukushima nuclear accident after the earthquake and tsunami of March 11, 2011, by five newspapers and the official internet sites of the Federal Service for Surveillance on Consumer Rights Protection and Human Well-being and the Federal Service for Hydrometeorology and Environmental Monitoring. The analysis of the data suggests that the population is satisfied with its request in timely, clear, and reliable information on the radiation situation, protective measures and forecast of the situation development in the future. The possibility to learn about different viewpoints positively affects the mood of readers and reduces their anxiety. However, it should be recognized that, if the information have been provided by a single media only, the population would not have adequate understanding of the situation related to the accident at the Fukushima nuclear power plant.

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

    Science.gov (United States)

    Kawaguchi, Takumi; Suetsugu, Takuro; Ogata, Shyou; Imanaga, Minami; Ishii, Kumiko; Esaki, Nao; Sugimoto, Masako; Otsuyama, Jyuri; Nagamatsu, Ayu; Taniguchi, Eitaro; Itou, Minoru; Oriishi, Tetsuharu; Iwasaki, Shoko; Miura, Hiroko; Torimura, Takuji

    2016-05-01

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

  17. THE ROLE OF BELARUS NATIONAL COMMISSION ON RADIATION PROTECTION IN THE MINIMIZATION OF CONSEQUENCES OF THE ACCIDENT AT THE CHERNOBYL NUCLEAR POWER PLANT

    Directory of Open Access Journals (Sweden)

    A. N. Stozharov

    2016-01-01

    Full Text Available The Belarus National Commission on Radiation Protection was established in 1991, based on the former Byelorussian Soviet Socialist Republic Supreme Council Resolution. The Commission works out recommendations on the radiation protection to submit to the state authorities, state institutions under the Republic of Belarus Government and state research institutions, reviews and assesses scientific data in the field of radiation protection and makes suggestions in regards of the implementation of the achieved developments. The Commission engages leading scientists and practitioners from Belarus, involved in the provision of the radiation protection and safety in the state. The methodological cornerstone for the Commission activities was chosen to be the committment to the worldwide accepted approach of the nature and magnitude of the undertaken protective measures justification in the field of radiation safety. The Commission adheres the ALARA optimization criteria as the core of the aforementioned approach. The Commission has also submited to the Government a number of developments which were crucial in the highest level managerial decisions elaboration. The latter impacted directly the state tactics and strategy in the environmental, health and social consequences of the Chernobyl disaster minimization. Following the recommendations of the international institutions (ICRP, IAEA, UNSCEAR, FAO/WHO, developments of the colleagues in the Russian Federation, Ukraine and the local regional experience, the Commission proceeds with the expert observation of the ongoing protective measures to reduce the radiation impact and population exposure resulted from the Chernobyl accident, is actively occupied in the radiation safety ensuring at the Belarussian nuclear power plant being under construction, much contributes to elaboration of the new version of the state Law “On Radiation Protection of Population” and other regulatory documents.

  18. Boolean Algebra Application in Analysis of Flight Accidents

    Directory of Open Access Journals (Sweden)

    Casandra Venera BALAN

    2015-12-01

    Full Text Available Fault tree analysis is a deductive approach for resolving an undesired event into its causes, identifying the causes of a failure and providing a framework for a qualitative and quantitative evaluation of the top event. An alternative approach to fault tree analysis methods calculus goes to logical expressions and it is based on a graphical representation of the data structure for a logic - based binary decision diagram representation. In this analysis, such sites will be reduced to a minimal size and arranged in the sense that the variables appear in the same order in each path. An event can be defined as a statement that can be true or false. Therefore, Boolean algebra rules allow restructuring of a Fault Tree into one equivalent to it, but simpler.

  19. The Factors Analysis on Food Safety Accidents Statistics

    Directory of Open Access Journals (Sweden)

    Boyi Xiang

    2015-05-01

    Full Text Available The study uses SPSS17.0 analysis of validity and reliability of the food enterprises questionnaire. Using AMOS17. 0 software for structural equation model test of goodness of fit and analysis of on the path. From the “melamine” to “Sudanred” and “steroid-tainted pork” events that have been exposed recently, series of typical food safety incidents resulted in the emergence of food safety issues become the focus of attention. A series of food processing can be contaminated by harmful substances, resulting in harmful food, thus constituting food safety issues and poses a serious threat to public and person’s health.

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

    Science.gov (United States)

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

    2013-10-01

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

  1. CFD Analysis of Migration Mechanism of Source Term Under Severe Accident

    Institute of Scientific and Technical Information of China (English)

    CHEN; Lin-lin; SUN; Xue-ting; JI; Song-tao

    2013-01-01

    The analysis of the migration of source term under severe accident is one of the important aspects of‘Studies on Migration Mechanism of the Source Term under Severe Accident’,which is a significant task of the National Large Advanced PWR Research Program.This research aims at building up a method for analyzing fission product behavior in the containment with CFD code.The effect of PCCS(Passive

  2. Exposure rate response analysis of criticality accident dectector at Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Zino, J.F.

    1995-01-01

    This analysis investigated the exposure response behavior of gamma-ray ionization chambers used in the criticality accident systems at the Savannah River Site (SRS). The project consisted of performing exposure response measurements with a calibrated {sup 137}Cs source for benchmarking of the MCNP Monte Carlo code. MCNP was then used to extrapolate the ion chamber`s response to gamma-rays with energies outside the current domain of measured data for criticality fission sources.

  3. Impact of traffic congestion on road accidents: a spatial analysis of the M25 motorway in England.

    Science.gov (United States)

    Wang, Chao; Quddus, Mohammed A; Ison, Stephen G

    2009-07-01

    Traffic congestion and road accidents are two external costs of transport and the reduction of their impacts is often one of the primary objectives for transport policy makers. The relationship between traffic congestion and road accidents however is not apparent and less studied. It is speculated that there may be an inverse relationship between traffic congestion and road accidents, and as such this poses a potential dilemma for transport policy makers. This study aims to explore the impact of traffic congestion on the frequency of road accidents using a spatial analysis approach, while controlling for other relevant factors that may affect road accidents. The M25 London orbital motorway, divided into 70 segments, was chosen to conduct this study and relevant data on road accidents, traffic and road characteristics were collected. A robust technique has been developed to map M25 accidents onto its segments. Since existing studies have often used a proxy to measure the level of congestion, this study has employed a precise congestion measurement. A series of Poisson based non-spatial (such as Poisson-lognormal and Poisson-gamma) and spatial (Poisson-lognormal with conditional autoregressive priors) models have been used to account for the effects of both heterogeneity and spatial correlation. The results suggest that traffic congestion has little or no impact on the frequency of road accidents on the M25 motorway. All other relevant factors have provided results consistent with existing studies.

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

    Science.gov (United States)

    Rao, Suman

    2007-04-11

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

  5. Radiological accidents: analysis of the information disseminated by media and public acceptance of nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Delgado, Jose Ubiratan; Tauhata, Luiz [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil); Garcia, Marcia Maria [Fundacao Inst. Oswaldo Cruz (FIOCRUZ), Rio de Janeiro, RJ (Brazil). Dept. de Virologia

    1995-12-31

    A methodology to treat quantitatively information by Media concerning a nuclear or a radiological accident is presented. It allows us to classify information according to the amount, importance and way of showing, into one indicator, named Information Equivalent. This establishes a procedure for analysis of released information and includes: number of head-lines, illustrations, printed lines, editorials, authorities quoted and so on. Interpretation becomes easier when the evolution and statistical trend of this indicator is observed. The application to evaluate the dissemination of the accident which took place in 1987 in Goiania, Brazil, was satisfactory and allowed us to purpose a model. This will aid the planning, the decision making process and it will improve relationships between technical staff and media during the emergency. (author). 5 refs., 4 figs., 3 tabs.

  6. Analysis of Early Severe Accident Initiated by LBLOCA for Qinshan Phase II Nuclear Power Project

    Directory of Open Access Journals (Sweden)

    Shi Xing-Wei

    2013-07-01

    Full Text Available The purpose of this study is to simulate an early Severe Accident (SA scenario more detail through transferring the thermal-hydraulic status of the plant predicted by RELAP5 computer code to SA Program (SAP. Based on the criterion of date extract time, the RELAP5 thermal-hydraulic calculation data is extracted to form a file for SAP input card at 1477K of cladding surface. Relying on the thermal-hydraulic boundary parameters calculated by RELAP5 code, analysis of early SA initiated by the Large Break Loss-of-Coolant Accident (LBLOCA without mitigation measures for Qinshan Phase II Nuclear Power Plant (QSP-II performed by SAP through finding the key events of accident sequence, estimating the amount of hydrogen generation and oxidation behavior of the cladding and evaluating the relocation order of the materials collapsed in the central region of the core. The results of this study are expected to improve the SA analysis methodology more detail through analyzing early SA scenario.

  7. Analysis and Simulation of Severe Accidents in a Steam Methane Reforming Plant

    Directory of Open Access Journals (Sweden)

    MohammadJavad Jafari

    2015-10-01

    Full Text Available Severe accidents of process industries in Iran have increased significantly in recent decade. This study quantitatively analyzes the hazards of severe accidents imposed on people, equipment and building by a hydrogen production facility. A hazard identification method was applied. Then a consequence simulation was carried out using PHAST 6.54 software package and at the end, consequence evaluation was carried out based on the best-known and different criteria. Most hazardous jet fire and flash fire will be occurred in desulfurization and reformer units respectively. The most dangerous vapor cloud explosion will be caused by a rupture in desorfurizing reactor. This incident with an overpressure of 0.83 bars at a distance of 45 m will kill all people and will destroy all buildings and equipments that are located at this distance. The safety distance determined by TNO Multi-Energy model and according to the worst consequence is equal to 260 m. Vapor cloud explosion will have the longest harmful distance on both human and equipment compared to jet fire and flash fire. Atmospheric condition will have a significant influence on harmful distance, especially in vapor cloud explosion. Therefore, the hydrogen production by natural  gas  reforming  is  a  high-risk  process  and  should  always  be  accompanied  by  the  full implementation of the safety rules, personal protection and equipment fireproofing and building blast proofing against jet fire and explosions.

  8. Questions concerning safety and risk after the nuclear accidents in Japan. Deepened accident analysis for the Fukushima Daiichi power plant; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Vertiefte Ereignisanalyse zur Anlage Fukushima-Daini

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2015-02-25

    The study questions concerning safety and risk in Japanese power plants following the disastrous nuclear accident covers the following issues: the nuclear facility Fukushima Daiichi, site characterization, important technical equipment, important electro-technical equipment, personal; description of the accident progression in the Fukushima nuclear power plant: impact of the earthquake, impact of the tsunami, short-term measures of the operating personnel, pressure and temperature situation in the containments, restoration of the after-heat cooling system in the units 1/2 and 4, fuel element storage pool, summarized parameters during the accident progress; comparative analysis of the accident progression at the Fukushima Daiichi site.

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

    Science.gov (United States)

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

    2015-05-01

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

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

    Science.gov (United States)

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

    2016-07-01

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

  11. Análise e classificação dos fatores humanos nos acidentes industriais Analysis and classification of the human factors in industrial accidents

    Directory of Open Access Journals (Sweden)

    Cármen Regina Pereira Correa

    2007-04-01

    Full Text Available O presente texto apresenta a evolução do conhecimento do fenômeno "acidente", mostrando a mudança do conceito do acidente como obra do destino para um componente do processo produtivo de qualquer segmento - industrial, aeronáutico, serviços, transporte dentre outros. O método de análise e classificação dos fatores humanos nos acidentes é apresentado e discutido quanto à viabilidade de implementação. Finalmente, conclui-se que a forma atual e moderna para prevenção de acidentes está baseada na identificação antecipada das falhas latentes da organização e do sistema, e que a ferramenta apresentada contribui para a gestão proativa e conseqüentemente para a diminuição do impacto dos acidentes do trabalho no processo produtivo.The present text presents the evolution of the knowledge of the phenomenon "accident", showing the change of the concept of the accident as workmanship of the destination for one component of the productive process of any segment - industrial, aeronautical, services, transports amongst others. The method of analysis and classification of the human factors in the accidents is presented and argued how much to the implementation viability. Finally one concludes that the current and modern form for prevention of accidents is based on the anticipated identification of the latent failures of the organization and the system, and that the presented tool contributes consequently for the pro-active management and in the reduction of the impact of the employment-related accidents in the productive process.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  14. Analysis of Accidents in Nine Iranian Gas Refineries: 2007–2011

    OpenAIRE

    AR Shakibmanesh; Bolouri, A.; R Mehrdad

    2013-01-01

    Background: Occupational accidents are one of the major health hazards in industries and associated with high mortality, morbidity, spiritual damage and economic losses in the world.Objective: To determine the incidence of occupational accidents in 9 Iranian gas refineries between March 2007 and February 2011.Methods: Data on all occupational accidents occurred between March 2007 and February 2011, as well as other possible associated variables including time of accident, whether the accident...

  15. Severe accident risks from external events

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

    This paper reviews the early development of design requirements for seismic events in USA early developing nuclear electric generating fleet.Notable safety studies,including WASH-1400,Sandia Siting Study and the NUREG-1150 probabilistic risk study,are briefly reviewed in terms of their relevance to extreme accidents arising from seismic and other severe accident initiators.Specific characteristic about the nature of severe accidents in nuclear power plant (NPP) are reviewed along with present day state-of-art analysis methodologies (methods for estimation of leakages and consequences of releases (MELCOR) and MELCOR accident consequence code system (MACCS)) that are used to evaluate severe accidents and to optimize mitigative and protective actions against such accidents.It is the aim of this paper to make nuclear operating nations aware of the risks that accompany a much needed energy resource and to identify some of the tools,techniques and landmark safety studies that serve to make the technology safer and to maintain vigilance and adequate safety culture for the responsible management of this valuable but unforgiving technology.

  16. Analysis of 78 cases of prehospital death due to traffic accident injury

    Institute of Scientific and Technical Information of China (English)

    胡孝菽; 洪勇; 等

    1999-01-01

    Objective The cause and time of prehospital death for the injured patients caused by traffic accidents were studied in order to improve traffic management and clinical treatment,and reduce mortality.Methods The characteristics of the injury,the rescue procedure,the status of the injury leading to death were analyzed based on the retrospective data of 78 cases died before admission.Results The main causes of prehospital death in the traffic accidents included:1.head injury,2.bleeding,3.chest and heart wound,4.spinal cord injury at upper cervix.Death happened immediately after injury was in 17 cases.Death happened from the accident site to our hospital was in 47 cases.Death happened within half an hour after reaching emergency room was in 14 cases.In all of the cases,the death on the transfer took up 62.5%.Conclusions Findings from analysis of the data will be presented on a wide range of traffic safety issues.These include enhancing education of traffic safety and administration of drivers and motor vehicles,establishing a perfect emergency medical service system and a well-trained team of first aid,and popularizing first aid knowledge to all people.

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

    Science.gov (United States)

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

    2016-01-01

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

  18. Road accidents at night in the Netherlands : a national analysis according to official road accident data. Contribution to OECD Research Group TS 3 on Improving Road Safety at Night.

    NARCIS (Netherlands)

    Harris, S.

    1979-01-01

    The questionnaire about night-time accident data of the OECD Research Group TS 3 on Improving Road Safety at Night was filled in for the Netherlands. Thereafter a national analysis was written, using the already completed accident data questionnaire. Guidelines for the contents and presentation fo

  19. Integration of human reliability analysis into the high consequence process

    Energy Technology Data Exchange (ETDEWEB)

    Houghton, F.K.; Morzinski, J.

    1998-12-01

    When performing a hazards analysis (HA) for a high consequence process, human error often plays a significant role in the hazards analysis. In order to integrate human error into the hazards analysis, a human reliability analysis (HRA) is performed. Human reliability is the probability that a person will correctly perform a system-required activity in a required time period and will perform no extraneous activity that will affect the correct performance. Even though human error is a very complex subject that can only approximately be addressed in risk assessment, an attempt must be made to estimate the effect of human errors. The HRA provides data that can be incorporated in the hazard analysis event. This paper will discuss the integration of HRA into a HA for the disassembly of a high explosive component. The process was designed to use a retaining fixture to hold the high explosive in place during a rotation of the component. This tool was designed as a redundant safety feature to help prevent a drop of the explosive. This paper will use the retaining fixture to demonstrate the following HRA methodology`s phases. The first phase is to perform a task analysis. The second phase is the identification of the potential human, both cognitive and psychomotor, functions performed by the worker. During the last phase the human errors are quantified. In reality, the HRA process is an iterative process in which the stages overlap and information gathered in one stage may be used to refine a previous stage. The rationale for the decision to use or not use the retaining fixture and the role the HRA played in the decision will be discussed.

  20. Consequence analysis in LPG installation using an integrated computer package.

    Science.gov (United States)

    Ditali, S; Colombi, M; Moreschini, G; Senni, S

    2000-01-07

    This paper presents the prototype of the computer code, Atlantide, developed to assess the consequences associated with accidental events that can occur in a LPG storage plant. The characteristic of Atlantide is to be simple enough but at the same time adequate to cope with consequence analysis as required by Italian legislation in fulfilling the Seveso Directive. The application of Atlantide is appropriate for LPG storage/transferring installations. The models and correlations implemented in the code are relevant to flashing liquid releases, heavy gas dispersion and other typical phenomena such as BLEVE/Fireball. The computer code allows, on the basis of the operating/design characteristics, the study of the relevant accidental events from the evaluation of the release rate (liquid, gaseous and two-phase) in the unit involved, to the analysis of the subsequent evaporation and dispersion, up to the assessment of the final phenomena of fire and explosion. This is done taking as reference simplified Event Trees which describe the evolution of accidental scenarios, taking into account the most likely meteorological conditions, the different release situations and other features typical of a LPG installation. The limited input data required and the automatic linking between the single models, that are activated in a defined sequence, depending on the accidental event selected, minimize both the time required for the risk analysis and the possibility of errors. Models and equations implemented in Atlantide have been selected from public literature or in-house developed software and tailored with the aim to be easy to use and fast to run but, nevertheless, able to provide realistic simulation of the accidental event as well as reliable results, in terms of physical effects and hazardous areas. The results have been compared with those of other internationally recognized codes and with the criteria adopted by Italian authorities to verify the Safety Reports for LPG

  1. Fluid-structure interaction analysis of a hypothetical core disruptive accident in LMFBRs

    Energy Technology Data Exchange (ETDEWEB)

    Liu Chuang [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China)]. E-mail: lch98@mails.tsinghua.edu.cn; Zhang Xiong [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Lu Mingwan [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China)

    2005-03-01

    To ensure safety, it is necessary to assess the integrity of a reactor vessel of liquid-metal fast breeder reactor (LMFBR) under HCDA. Several important problems for a fluid-structural interaction analysis of HCDA are discussed in the present paper. Various loading models of hypothetical core disruptive accident (HCDA) are compared and the polytropic processes of idea gas (PPIG) law is recommended. In order to define a limited total energy release, a '5% truncation criterion' is suggested. The relationship of initial pressure of gas bubble and the total energy release is given. To track the moving interfaces and to avoid the severe mesh distortion an arbitrary Lagrangrian-Eulerian (ALE) approach is adopted in the finite element modeling (FEM) analysis. Liquid separation and splash from a free surface are discussed. By using an elasticity solution under locally uniform pressure, two simplified analytical solutions for 3D and axi-symmetric case of the liquid impact pressure on roof slab are derived. An axi-symmetric finite elements code FRHCDA for fluid-structure interaction analysis of hypothetical core disruptive accident in LMFBR is developed. The CONT benchmark problem is calculated. The numerical results agree well with those from published papers.

  2. Analysis of credible accidents for Argonaut reactors. Report for October 1980-April 1981

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: insertion of excess reactivity, catastrophic rearrangement of the core, explosive chemical reaction, graphite fire, and a fuel-handling accident.

  3. Development of system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1992-01-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.

  4. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  5. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  6. Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

    Directory of Open Access Journals (Sweden)

    Kil-Mo Koo

    2012-01-01

    Full Text Available Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard requires that the normal signal level for pressure, flow, and resistance temperature detector sensors be in the range of 4~20 mA for most instruments. Whereas, in the case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response analysis have been evaluated with an equivalent circuit model of real instruments, which is regarded as the best method. The main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for equivalent circuit modeling. The program links signal analysis tool code to abnormal signal simulation engine code not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis module developed to obtain a varying range of the R-C circuit elements in high temperature conditions. As a result, a special function for abnormal pulse signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output pulse signals through a response characteristic of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

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

    OpenAIRE

    Mehran Amiri; Abdollah Ardeshir; Mohammad Hossein Fazel Zarandi

    2014-01-01

    Abstract Background The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. Methods The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were d...

  8. Utilization of the IAIA (Investigation and Analysis of Incidents and Accidents) method in the investigation of the P-36 platform accident; Utilizacao do metodo IAIA (Investigacao e Analise de Acidentes e Incidentes) na investigacao do acidente ocorrido na plataforma P-36

    Energy Technology Data Exchange (ETDEWEB)

    Teles, Marcus de Barros [ARCE - Agencia Reguladora de Servicos Publicos Delegados do Estado do Ceara, Fortaleza, CE (Brazil)

    2004-07-01

    In the beginning of XXI century the Brazilian oil industry report a big accident involving that which was the biggest petroleum platform of the world. With capacity production of 180.000 barrels a day and capacity compression of 7,2 million cubic meter a day of natural gas, the off-shore platform P-36 was situated on Roncador field, in Campos basin, operating in 1360 meters of water. As consequences, eleven deaths with irreparable traumas to the families, friends and worker partners, one billion dollars in prejudices to brazilian country, environmental damages by oil leak and injuries to PETROBRAS reputation in Brazil and in the world. The method of investigation and analysis of incidents and accidents - IAIA is very wide and its philosophy contain a lot of topics, since basic concepts, investigation actions, analysis action and diagnosis by the general kind of fail. Using this method and taking advantage from the report elaborated by the commission organized by ANP - Agencia Nacional do Petroleo and DPC - Diretoria de Portos e Costas, responsible for the investigation and analysis of the accident occurred with P-36, this paper identify the direct and indirect causes of the accident, in attempt to avoid new similar situations. (author)

  9. Accidents with sulfuric acid

    OpenAIRE

    Rajković Miloš B.

    2006-01-01

    Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eigh...

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

    Science.gov (United States)

    Stephens, Mitchell; Edison, Nadyne G.

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

  11. Analysis of Public Perception on Radiation: with One Year after Fukushima Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bang Ju [Korean Science Reporters Association, Seoul (Korea, Republic of)

    2012-03-15

    A year has passed since the nuclear power plant accident in Fukushima on March 11, 2011, and a survey for public perception on radiation by Korean people has been made. The methodological design was based on a quantitative survey and a frequency analysis was done. The analysis objects were survey papers (n=2,754pcs) answered by random ordinary citizens chosen from all over the country. The questionnaires, and study tool, were directly distributed and collected. A total of 40 questionnaires using a coefficient of Cronbach's {alpha} per each area was 'self perception of radiation' (0.620), 'radiation risk' (0.830), 'benefit from radiation' (0.781), 'radiation controlled' (0.685), 'informative source of radiation' (0.831), 'influence degree from Fukushima accident' (0.763), showing rather high score from all areas. As the result of the questionnaires, the knowledge of radiation concept was 69.50 out of 100 points, which shows a rather significant difference from the result of 'know well about radiation' (53.7%) and 'just know about radiation' (37.40%). According to the survey, one of the main reasons why radiation seems risky was that once exposed to radiation, it may not have negative impacts presently but, the next generation could see negative impacts (66.1%). About 41% of our respondents showed a negative position against the government's report on radiation while 39.5% of respondents said that we should stop running nuclear power in light of Fukushima nuclear power plant accident. This study was done for the first time by Korean people's public perception on radiation after the Fukushima nuclear power plant accident. We expect this might have significant contributions to the establishment of the government's policy on radiation.

  12. Analysis and Consequences of the Iridium 33-Cosmos 2251 Collision

    Science.gov (United States)

    Anz-Meador, P. D.; Liou, Jer-Chi

    2010-01-01

    The collision of Iridium 33 and Cosmos 2251, on 10 February 2009, was the first known unintentional hypervelocity collision in space of intact satellites. Iridium 33 was an active commercial telecommunications satellite, while Cosmos 2251 was a derelict communication satellite of the Strela-2M class. The collision occurred at a relative velocity of 11.6 km/s at an altitude of approximately 790 km over the Great Siberian Plain and near the northern apex of Cosmos 2251 s orbit. This paper describes the physical and orbital characteristics of the relevant spacecraft classes and reports upon our analysis of the resulting debris clouds size, mass, area-to-mass ratio, and relative velocity/directionality distributions. We compare these distributions to those predicted by the NASA breakup model and notable recent fragmentation events; in particular, we compare the area-to-mass ratio distribution for each spacecraft to that exhibited by the FY-1C debris cloud for the purpose of assessing the relative contribution of modern aerospace materials to debris clouds resulting from energetic collisions. In addition, we examine the long-term consequences of this event for the low Earth orbit (LEO) environment. Finally, we discuss "lessons learned", which may be incorporated into NASA s environmental models.

  13. Proceeding of the workshop on the results of the cooperative research between JAERI and CHESCIR concerning the study on assessment and analysis of environmental radiological consequences and verification of an assessment system

    Energy Technology Data Exchange (ETDEWEB)

    Amano, Hikaru; Saito, Kimiaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    This workshop was organized and sponsored by the Japan Atomic Energy Research Institute (JAERI) and Chernobyl Science and Technology Center for International Research (CHESCIR). JAERI and CHESCIR have conducted 8 years research cooperation from 1992 to 1999 concerning the study on assessment and analysis of environmental radiological consequences and verification of an assessment system, focusing on the Chernobyl contaminated area. It contained 3 research subjects. Subject-1 initiated at 1992 and focused the study on measurements and evaluation of environmental external exposure after nuclear accident. Subject-2 initiated at 1992 and focused the study on the validation of assessment models in an environmental consequence assessment methodology for nuclear accidents. Subject-3 initiated at 1995 and focused on the study on migration of radionuclides released into terrestrial and aquatic environment after nuclear accidents. This workshop was held to summarize the research cooperation between JAERI and CHESCIR, and to discuss future research needs in this field. (author)

  14. Accident Process and Consequence Research for LOCA Combining with Blackout Accident of Ship Reactor%船用堆破口叠加全船断电事故进程及后果研究

    Institute of Scientific and Technical Information of China (English)

    张帆; 陈航; 张彦招; 晏峰

    2015-01-01

    Using MELCOR code ,the combination of LOCA and blackout accident of ship reactor was modeled and calculated , and the accident process and source term release were researched . The results show that the accident leads to lower head of pressure vessel and bilge creep‐rupture finally without emergency power .The release fraction of inert gases and iodine are above 80% ,the main form of iodine is CsI with great deposit and less airborne fraction .The accident process is decided by the equiva‐lent diameter of break size .The production of H2 is decided by core temperature and water remaining in the core ,but has nothing to do with equivalent diameter of break size .T he probability of H2 detonation is unlikely to occur .T he results can provide tech‐nical support for emergency maintenance and emergency decision‐making .%采用M ELCOR程序,对船用堆破口叠加全船断电事故进行建模计算,并对事故进程和源项释放进行了研究。计算结果表明:若应急电源无法投入,最终将导致压力容器下封头失效和舱底失效;所研究事故的惰性气体、碘释放量均在80%以上,且释放的I主要以CsI形式存在,滞留量大,气载量小。事故进展快慢取决于破口当量尺寸,但氢气的产量与堆芯温度、堆芯残余水量相关,与破口当量尺寸无直接关系,堆舱内发生氢爆可能性不大。本文计算结果可为应急抢修和应急决策提供技术支持。

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

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

    2007-04-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-09-01

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

  17. Analysis of hot leg natural circulation under station blackout severe accident

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg, and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg.The recirculation ratio and the hot mixing factor are also calculated and discussed.

  18. IMMEDIATE MENTAL CONSEQUENCES OF THE GREAT EAST JAPAN EARTHQUAKE AND FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT ON MOTHERS EXPERIENCING MISCARRIAGE, ABORTION, AND STILLBIRTH: THE FUKUSHIMA HEALTH MANAGEMENT SURVEY

    OpenAIRE

    Yoshida-Komiya, Hiromi; Goto, Aya; Yasumura, Seiji; FUJIMORI, KEIYA; Abe, Masafumi; FOR THE PREGNANCY AND BIRTH SURVEY GROUP OF THE FUKUSHIMA HEALTH MANAGEMENT SURVEY,

    2015-01-01

    Background: The Fukushima Pregnancy and Birth Survey was launched to monitor pregnant mothers’ health after the Great East Japan Earthquake and Fukushima Daiichi Nuclear Power Plant (NPP) accident. Several lines of investigations have indicated that a disaster impacts maternal mental health with childbirth. However, there is no research regarding mental health of mothers with fetal loss after a disaster. In this report, we focus on those women immediately after the Great East Japan Earthquake...

  19. Immediate mental consequences of the great east Japan earthquake and Fukushima nuclear power Plant accident on mothers experiencing miscarriage, abortion, and stillbirth: the Fukushima health management survey

    OpenAIRE

    YOSHIDA-KOMIYA, HIROMI; Goto, Aya; Yasumura, Seiji; FUJIMORI, KEIYA; Abe, Masafumi

    2015-01-01

    Background: The Fukushima Pregnancy and Birth Survey was launched to monitor pregnant mothers' health after the Great East Japan Earthquake and Fukushima Daiichi Nuclear Power Plant (NPP) accident. Several lines of investigations have indicated that a disaster impacts maternal mental health with childbirth. However, there is no research regarding mental health of mothers with fetal loss after a disaster. In this report, we focus on those women immediately after the Great East Japan Earthquake...

  20. Accident Analysis and Countermeasures of the Enterprises Involved in Ammonia%涉氨企业事故分析与对策

    Institute of Scientific and Technical Information of China (English)

    卢均臣; 王延平

    2015-01-01

    从事故发生环节、事故类型、事故设备、事故原因等几个方面分析了2005年-2014年全国涉氨企业发生的事故。分析表明:储存和使用环节事故最多;事故主要类型事故泄漏和中毒;主要发生在食品厂、肉类加工厂、冷饮厂、水产公司、果蔬公司、制药厂的制冷车间;主要发生在管道、储罐、阀门、法兰等部位;材料失效事故占比最高。最后,提出了预防此类事故的建议措施。%Accidents occurred in 2005-2014 in our country were analyzed from the aspects of the accident link, accident type, accident equipment, accident cause. Analysis showed that accidents in links of the storage and use were the most, the main types of accidents were leakage and poisoning, accidents mainly occured in food factory, meat processing factory, beverage factory, etc, accidents mainly occured in pipeline, storage tanks, valves, flanges, etc, the amount of accidents caused by material failure was the largest. Suggestions for preventing such accidents were put forward.

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

    CERN Document Server

    Wheatley, Spencer; Sornette, Didier

    2015-01-01

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

  2. The Chernobyl accident 20 years on: an assessment of the health consequences and the international response O acidente de Chernobyl 20 anos depois: avaliação das conseqüências e resposta internacional

    Directory of Open Access Journals (Sweden)

    Keith Baverstock

    2007-06-01

    Full Text Available Twenty years after the Chernobyl accident the WHO and the International Atomic Energy Authority issued a reassuring statement about the consequences. Our objectives in this study were to evaluate the health impact of the Chernobyl accident, assess the international response to the accident, and consider how to improve responses to future accidents. So far, radiation to the thyroid from radioisotopes of iodine has caused several thousand cases of thyroid cancer but very few deaths; exposed children were most susceptible. The focus on thyroid cancer has diverted attention from possible nonthyroid effects. The international response to the accident was inadequate and uncoordinated, and has been unjustifiably reassuring. Accurate assessment in future health effects is not currently possible in the light of dose uncertainties, current debates over radiation actions, and the lessons from the late consequences of atomic bomb exposure. Because of the uncertainties from and the consequences of the accident, it is essential that investigations of its effects should be broadened and supported for the long term. The United Nations should initiate an independent review of the actions and assignments of the agencies concerned, with recommendations for dealing with future international-scale accidents. These should involve independent scientists and ensure cooperation rather than rivalry.Vinte anos após o acidente de Chernobyl ocorrido em 1986, a OMS e a Autoridade Internacional sobre Energia Atômica lançaram um relatório sobre as conseqüências desse desastre. Nosso objetivo neste estudo é avaliar o impacto de tal acidente sobre a saúde e a reação internacional sobre o ocorrido, além de considerar se é possível melhorar as respostas em futuros desastres. Observamos que a radiação sobre a tireóide, proveniente de radioisótopos de iodo, causou milhares de casos de câncer, mas poucas mortes; as crianças expostas foram as mais suscetíveis. O

  3. A comparison of the hazard perception ability of accident-involved and accident-free motorcycle riders.

    Science.gov (United States)

    Cheng, Andy S K; Ng, Terry C K; Lee, Hoe C

    2011-07-01

    Hazard perception is the ability to read the road and is closely related to involvement in traffic accidents. It consists of both cognitive and behavioral components. Within the cognitive component, visual attention is an important function of driving whereas driving behavior, which represents the behavioral component, can affect the hazard perception of the driver. Motorcycle riders are the most vulnerable types of road user. The primary purpose of this study was to deepen our understanding of the correlation of different subtypes of visual attention and driving violation behaviors and their effect on hazard perception between accident-free and accident-involved motorcycle riders. Sixty-three accident-free and 46 accident-involved motorcycle riders undertook four neuropsychological tests of attention (Digit Vigilance Test, Color Trails Test-1, Color Trails Test-2, and Symbol Digit Modalities Test), filled out the Chinese Motorcycle Rider Driving Violation (CMRDV) Questionnaire, and viewed a road-user-based hazard situation with an eye-tracking system to record the response latencies to potentially dangerous traffic situations. The results showed that both the divided and selective attention of accident-involved motorcycle riders were significantly inferior to those of accident-free motorcycle riders, and that accident-involved riders exhibited significantly higher driving violation behaviors and took longer to identify hazardous situations compared to their accident-free counterparts. However, the results of the regression analysis showed that aggressive driving violation CMRDV score significantly predicted hazard perception and accident involvement of motorcycle riders. Given that all participants were mature and experienced motorcycle riders, the most plausible explanation for the differences between them is their driving style (influenced by an undesirable driving attitude), rather than skill deficits per se. The present study points to the importance of

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

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2012-01-01

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

  5. Analysis of a hypothetical loss of coolant accident in a Konvoi type NPP by GASFLOW and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Stefan; Royl, Peter [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Band, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    The 3D computational fluid dynamics code GASFLOW and the German containment code system COCOSYS, which is based on a lumped-parameter approach, are used to simulate the hydrogen-air-steam distribution and hydrogen mitigation in a Konvoi type nuclear power plant in a postulated hypothetical core melt accident. A break in a coolant loop and the subsequent loss of the coolant causes a strong heat-up of the core. As a consequence hydrogen is produced by oxidation of cladding tubes. The residual steam and the produced hydrogen are released into the containment through the break in the coolant loop. Without suitable counter measures, sensitive mixtures can build up with a combustion potential which could threaten the integrity of the containment. A model of a Konvoi type nuclear power plant which is equipped with passive autocatalytic recombiners is used to simulate such accident scenario. COCOSYS allows comprehensive simulation of all relevant processes of severe accidents, whereas GASFLOW is primarily designed to simulate the distribution of steam and hydrogen within the containment. This paper presents the comparison of GASFLOW and COCOSYS simulation results for the in-vessel phase of the selected accident. (orig.)

  6. Effect of Check Valve on Consequences of Coolant Pump Rotor Seizure Accident for EPR Reactor%止回阀对EPR反应堆主泵卡轴事故后果的影响

    Institute of Scientific and Technical Information of China (English)

    陈秋炀; 周拥辉

    2012-01-01

    分析计算欧洲先进压水堆(EPR)反应堆主泵卡轴事故,并对比在主泵出口安装止回阀和没有安装止回阀模型的卡轴事故安全分析.结果表明,在EPR主泵卡轴事故中,止回阀可增加模型堆芯进口流量约4%,有利于堆芯的冷却.止回阀可显著地提高堆芯最小偏离泡核沸腾比(DNBR),降低堆芯偏离泡核沸腾(DNB)份额,降低包壳温度约14℃.模型分析结果表明,在主泵卡轴事故工况下,主泵出口安装止回阀可更好地维持堆芯的完整性.%Counter current flow phenomenon would appear during reactor coolant pump rotor seizure accident. Present work analyzes the coolant pump rotor seizure accident for European Pressurized Reactor (EPR). The accident safety analysis results of model with check valve and without check valve are compared. It can be found that the check valve can increase the core inlet flow rate of model about 4%. The increasing of coolant flow rate is beneficial to the reactor core cooling. Check valve can increase the minimum departure from nucleate boiling ratio (DNBR), reduce the departure from nucleate boiling (DNB) fraction and the fuel rod cladding temperature about 14℃ during coolant pump rotor seizure accident. The analyses results show that the model with check valve can maintain the integrity of nuclear fuel rod effectively during reactor coolant pump rotor seizure accident.

  7. Analysis of traffic accident size for Korean highway using structural equation models.

    Science.gov (United States)

    Lee, Ju-Yeon; Chung, Jin-Hyuk; Son, Bongsoo

    2008-11-01

    Accident size can be expressed as the number of involved vehicles, the number of damaged vehicles, the number of deaths and/or the number of injured. Accident size is the one of the important indices to measure the level of safety of transportation facilities. Factors such as road geometric condition, driver characteristic and vehicle type may be related to traffic accident size. However, all these factors interact in complicate ways so that the interrelationships among the variables are not easily identified. A structural equation model is adopted to capture the complex relationships among variables because the model can handle complex relationships among endogenous and exogenous variables simultaneously and furthermore it can include latent variables in the model. In this study, we use 2649 accident data occurred on highways in Korea and estimate relationship among exogenous factors and traffic accident size. The model suggests that road factors, driver factors and environment factors are strongly related to the accident size.

  8. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

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

    Science.gov (United States)

    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

  10. A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Daeil; Kim, J. H.; Jang, S. C

    2007-03-15

    Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post-accident

  11. Structure of the thyroid pathology in the radiation exposed areas of Leningrad region: late consequences of Chernobyl accident after 20 years

    Energy Technology Data Exchange (ETDEWEB)

    Semenov, A.; Uspenskaya, A.; Bychenkova, E.; Chinchuk, I.; Novokshonov, K.; Chernikov, R.; Sleptsov, I.; Bubnov, A.; Fedotov, Y.; Makarin, V.; Karelina, Y. [Endocrinology, NWRMC FHSDA, ST-Petersburg (Russian Federation)

    2012-07-01

    After the Chernobyl accident large areas of the USSR were contaminated with fallout, it has been proved that I{sup 131} caused higher incidence of papillary thyroid cancer in children and adolescents. Further observation for over 20 years showed retention of high annual prevalence of this pathology among the population. The aim of this study is to evaluate the ultimate result of the influence of I{sup 131} on the thyroid gland. The study included 454 women living in localities affected by the Chernobyl accident in April-May 1986 (case) and 909 women living in fallout-free localities (ICCIDD method). The incidence of malignant thyroid tumors among the inhabitants of the contaminated territories is higher than in the control area. This phenomenon can not be unambiguously attributed to radiation induced cancers, but requires further investigation, perhaps by the method of carrying out continuous and all-round prophylactic medical examination. High incidence of autoimmune changes can be considered to have been caused by the action of I{sup 131} and prophylactic supplement with stable iodine

  12. An analysis on the severe accident progression with operator recovery actions

    Energy Technology Data Exchange (ETDEWEB)

    Vo, T.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Song, J.H., E-mail: dosa@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, T.W.; Kim, D.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of)

    2014-12-15

    Highlights: • Severe accident progression for the station blackout and SBLOCA accident. • Analyses on APR1400 using MELCOR. • Operator recovery actions for decay heat removal and inventory make up. • Determine the time allowed for the operator to prevent reactor vessel failure. • Insight for the operator recovery actions for the severe accident management. - Abstract: Analyses on the severe accident progressions for the station blackout (SBO) accident and small break LOCA (SBLOCA) initiated severe accident were performed for APR1400 by using MELCOR computer code. Operator recovery actions for decay heat removal and inventory make up using a depressurization system and safety injection pump were simulated in parallel with a simulation of the severe accident progression. Sensitivity studies on the operator actions were performed to investigate the changes in the timing of the reactor vessel failure and to determine the time allowed for the operator to prevent reactor vessel failure. Sensitivity analyses on the effect of major modeling parameters were performed additionally to quantify the uncertainties in timing. It is found that the operator has about 2 h for the recovery actions after the indication of core damage by the signal of core exit thermocouple (CET) for the SBLOCA initiated severe accident, while the operator has to take immediate actions after the indication of core damage by CET for the SBO accident.

  13. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  14. Learning from incidents and accidents

    NARCIS (Netherlands)

    Drupsteen, L.; Kampen, J. van

    2014-01-01

    There are many different definitions for what constitutes an incident or an accident, however the focus is always on unintended and often unforeseen events that cause unintended consequences. This article is focused on the process of learning from incidents and accidents. The focus is on making sure

  15. A Review of Criticality Accidents 2000 Revision

    Energy Technology Data Exchange (ETDEWEB)

    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. Safety analysis of a loss-of-coolant accident in a breeding blanket for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rocco, P.; Casini, G.; Djerassi, H.; Papa, L.; Pautasso, G.; Renda, V.; Rouyer, J.L.

    1985-07-01

    A LOCA in a blanket design proposed for NET (Next European Torus) is investigated. The structural analysis of a damaged breeder unit shows that this first containment barrier has a high probability of survival to this accident. The radioactive sources involved are evaluated and an assessment is made of all containment barriers and associated protection systems.

  17. SisRadiologia: a new software tool for analysis of radiological accidents and incidents in industrial radiography

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Camila M. Araujo; Silva, Francisco C.A. da, E-mail: araujocamila@yahoo.com.br, E-mail: dasilva@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Araujo, Rilton A., E-mail: consultoria@maximindustrial.com.br [Maxim Industrial Assessoria TI, Rio de Janeiro, RJ (Brazil)

    2013-07-01

    According to the International Atomic Energy Agency (IAEA), many efforts have been made by Member states, aiming a better control of radioactive sources. Accidents mostly happened in practices named as high radiological risk and classified by IAEA in categories 1 and 2, being highlighted those related to radiotherapy, large irradiators and industrial radiography. Worldwide, more than 40 radiological accidents have been recorded in the industrial radiography. Worldwide, more than 40 radiological accidents have been recorded in the industrial radiography area, involving 37 workers, 110 members of the public and 12 fatalities. Records display 5 severe radiological accidents in industrial radiography activities in Brazil, in which 7 workers and 19 members of the public were involved. Such events led to hands and fingers radiodermatitis, but to no death occurrence. The purpose of this study is to present a computational program that allows the data acquisition and recording in the company, in such a way to ease a further detailed analysis of radiological event, besides providing the learning cornerstones aiming the avoidance of future occurrences. After one year of the 'Industrial SisRadiologia' computational program application - and mostly based upon the workshop about Analysis and Dose Calculation of Radiological Accidents in Industrial Radiography (Workshop sobre Analise e Calculo de dose de acidentes Radiologicos em Radiografia Industrial - IRD 2012), in which several Radiation Protection officers took part - it can be concluded that the computational program is a powerful tool to data acquisition, as well as, to accidents and incidents events recording and surveying in Industrial Radiography. The program proved to be efficient in the report elaboration to the Brazilian Regulatory Authority, and very useful in workers training to fix the lessons learned from radiological events.

  18. Analysis of the consequences of aircraft manufacturers’ system integration model

    Directory of Open Access Journals (Sweden)

    João Henrique Lopes Guerra

    2013-11-01

    Full Text Available This is a theoretical-conceptual, which aimed to identify some likely consequences of the integration model systems that have been adopted in the aerospace industry by major aircraft manufacturers in the world. In the model of system integration, these manufacturers maintain internally the activities associated with their basic skills and transfer their skills to peripheral suppliers. We identified the following consequences: the growth of strategic alliances in the airline industry, the internationalization of aeronautical chains, with the strengthening of productive activities in some geographic regions; challenges related to the domestic supplier base and the consolidation of national chains, the greatest power suppliers of the first layer, the contribution to the dissemination of knowledge among supply chains, and the potential emergence of new competitors.

  19. Pattern extraction for high-risk accidents in the construction industry: a data-mining approach.

    Science.gov (United States)

    Amiri, Mehran; Ardeshir, Abdollah; Fazel Zarandi, Mohammad Hossein; Soltanaghaei, Elahe

    2016-09-01

    Accidents involving falls and falling objects (group I) are highly frequent accidents in the construction industry. While being hit by a vehicle, electric shock, collapse in the excavation and fire or explosion accidents (group II) are much less frequent, they make up a considerable proportion of severe accidents. In this study, multiple-correspondence analysis, decision tree, ensembles of decision tree and association rules methods are employed to analyse a database of construction accidents throughout Iran between 2007 and 2011. The findings indicate that in group I, there is a significant correspondence among these variables: time of accident, place of accident, body part affected, final consequence of accident and lost workdays. Moreover, the frequency of accidents in the night shift is less than others, and the frequency of injury to the head, back, spine and limbs are more. In group II, the variables time of accident and body part affected are mostly related and the frequency of accidents among married and older workers is more than single and young workers. There was a higher frequency in the evening, night shifts and weekends. The results of this study are totally in line with the previous research.

  20. A Study for Appropriateness of National Nuclear Policy by using Economic Analysis Methodology after Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jong Myoung; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    The aim of this paper is to clarify the appropriateness of national nuclear policy in BPE of Korea from an economic perspective. To do this, this paper only focus on the economic analysis methodology without any considering other conditions such as political, cultural, or historical things. In a number of countries, especially Korea, nuclear energy policy is keeping the status quo after Fukushima accident. However the nation's nuclear policy may vary depending on the choice of people. Thus, to make the right decisions, it is important to deliver accurate information and knowledge about nuclear energy to the people. As proven in this paper, the levelized cost of nuclear power is the most inexpensive among the base load units. As the reliance on nuclear power is getting stronger through the economic logic, the nuclear safety and environmental elements will be strengthened. Based on this, national nuclear policy should be promoted. In the aftermath of the Fukushima accident recognized as the world's worst nuclear disaster since the Chernobyl, there are some changes in the nuclear energy policy of various countries. Germany, for example, called a halt to operate Nuclear Power Plant (NPP) which accounts for about 7.5% of the national power generation capacity of 6.3GW. In developing countries such as China and India they conducted the safety check of the nuclear power plants again before preceding their nuclear business. Korea government announced 'The 6th Basic Plan for Long-term Electricity Supply and Demand (BPE)', considering the safety and general public acceptance of the nuclear power plants. According to BPE, they postponed a plan for additional NPP construction, except for constructions that had been already reflected in the 5th BPE. All told, the responses for nuclear energy policy of countries are different depending on their own circumstances.

  1. Serious work accidents and their causes - An analysis of data from Eurostat

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2015-01-01

    in Europe each year. Despite the uncertainty of the data collected by Eurostat over two years stile provide a picture of the seriousness of the accidents, the sources of risk and the events taking place when the accidents occur. Data from Eurostat were analysed to find out which hazards and accidental...

  2. Report on the consequences of Chernobylsk accident in France Minister missions from the 25. february to 6. august 2002; Rapport sur les consequences de l'accident de Tchernobyl en France missions ministerielles du 25 fevrier et du 6 aout 2002

    Energy Technology Data Exchange (ETDEWEB)

    Aurengo, A

    2006-04-15

    Actually, we have not any map that gives reliable quantitative data of Chernobylsk accident fallout on soils. The maps proposed for these deposits give order of magnitude; they find east-west gradient conform to the origin of the accident and confirm the importance of the rain. But the quantitative value is only an approximation where the precision is not known (error interval). It does not allow to know the radiation doses to the thyroid because the food contamination does not increase like the soils contamination. It could be possible to improve the models but the scientific council of I.R.S.N. proposes to realize a periodic ground state of soils contamination in cesium. It would be a better step of a more reliable mapping of Chernobylsk accident fallout. (N.C.)

  3. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Science.gov (United States)

    2010-07-01

    ... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection...

  4. The dominance of accidents caused by banalities

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    Most prevention analysis is focused on high risks, such as explosion, fire, lack of containment for chemicals, crashes in transportation systems, lack of oxygen, or chemical poisoning. In the industrial world, these kinds of risk still lead to incidents with huge consequences, albeit very seldom...... as an example of how much information such systems can offer in general for the work of accident prevention in more traditional and common enterprises....

  5. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

    Information concerning 520 bicycle accidents and their victims was obtained from medical records and the victims' replies to questionnaires. The analyzed aspects included risk of injury, completeness of accident registrations by police and in hospitals, types of injuries and influence of the cyclists' age and sex, alcohol, fatigue, hunger, haste, physical disability, purpose of cycling, wearing of protective helmet and other clothing, type and quality of road surface, site of accident (road junctions, separate cycle paths, etc.) and turning manoeuvres.

  6. A Human Reliability Analysis of Pre-Accident Human Errors in the Low Power and Shutdown PSA of the KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Daeil; Jang, Seungchul

    2007-03-15

    Korea Atomic Energy Research Institute, using the ANS Low Power /Shutdown (LPSD)PRA Standard, evaluated the LPSD PSA model of the KSNP, Younggwang (YGN) Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the pre-accident human errors in the LPSD PSA model of the KSNP showed that 13 items among 15 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for pre-accident human errors in the LPSD PSA model for the KSNP to improve its quality. We considered potential pre-accident human errors for all manual valves and control/instrumentation equipment of the systems modeled in the KSNP LPSD PSA model except reactor protection system/ engineering safety features actuation system. We reviewed 160 manual valves and 56 control/instrumentation equipment. The number of newly identified pre-accident human errors is 101. Among them, the number of those related to testing/maintenance tasks is 56. The number of those related to calibration tasks is 45. The number of those related to only shutdown operation is 10. It was shown that the pre-accident human errors related to only shutdown operation contributions to the core damage frequency of LPSD PSA model for the KSNP was negligible.The self-assessment results for the new HRA results of pre-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II or III. It is expected that the HRA results for the pre-accident human errors presented in this study will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of supporting requirements for the postaccident human errors in the ANS LPSD PRA Standard.

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

    Science.gov (United States)

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

    2016-04-01

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

  8. Analysis on distribution of freeway accidents under various conditions in China

    Directory of Open Access Journals (Sweden)

    Xiaofei Wang

    2016-08-01

    Full Text Available This study aims to provide a current survey on the situation of freeway accidents in China. The results show that the accident rate, death toll, injury toll, and direct loss of property are 3.2, 8.4, 7.2, and 24.3 times that of the average for an ordinary highway in China. Freeway accidents occur mainly in Southern (20.77% and Central (20.2% China. With detailed data from Guangdong Province, the number of accidents in freeways with 80 km/h design speed (29.58/km/103 pcu was more than three times that in freeways with design speed of 100 km/h (9.54/km/103 pcu and 120 km/h (9.42/km/103 pcu. The total accident rate increased monotonously with the decrease in horizontal radius. The results indicate that 54.54% (/km/103 pcu of accidents occurred on a steep slope (4%–5%, representing about 10 times that of 3%–4% slope and 20 times that of the less than 3% slope. Based on the data, the safety situation of China’s highway transportation is obviously grim, and improving freeway heavy traffic management in economically developed regions, strengthening the safeguarding of mountain freeways, applying small radius and large vertical grade with caution, and developing a monitoring system of tunnels and interchanges could be used as effective measures to prevent freeway accidents.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  10. A Sensitivity Analysis of a Pipe Break Accident in a Preliminary Specific Design of the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Jeong, Jae Ho; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is a pool type sodium cooled fast reactor with a thermal power of 392.1 MW which has been developed in accord with an enhanced safety, an efficient utilization of uranium resources and a reduction of a high level waste volume in the Korea Atomic Energy Research Institute (KAERI) since 2012 under a National Nuclear R and D Program. The PGSFR has an inherent safety characteristic owing to the design to have a negative power reactivity coefficient during all operation modes and it has a passive safety characteristic due to the design of a passive decay heat removal circuit. In order to assess the inherent safety features of the PGSFR, a safety analysis was performed for a pipe break accident with MARS-LMR. And, the sensitivity studies were also performed to find the most conservative condition. As a result, the PGSFR was appropriately tripped by a high power to PHTS flow ratio using the method of extracting the PHTS flow rate from the pressure drop. The air flow rate was the most sensitive variable in the sensitivity analysis. Therefore, it is important to know the accurate uncertainty of the air flow rate in the AHX.

  11. Finite element based stress analysis of BWR internals exposed to accident loads

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E.; Weiss, F.P.; Werner, M.; Willschuetz, H.G.

    1998-10-01

    During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. Three dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core structure and lower core structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. (orig.)

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

    Science.gov (United States)

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

    2010-01-01

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

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

    Science.gov (United States)

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

    2013-01-01

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

  14. Safety analysis of 5 MW IEAR-1 reactor; Analise de seguranca do reator IEA-R1 a 5 MW

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Antonio T. e; Maprelian, Eduardo; Rodrigues, Antonio C.I.; Cabral, Eduardo L.L.; Molnary, Leslie de; Mesquita, Ricardo N.; Mendonca, Arlindo G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores. E-mail: teixeira@net.ipen.br

    2000-07-01

    This paper presents the methods and procedures utilized in the safety analysis of IEA-R1 research reactor. Four postulated accidents are quantitatively analyzed, being the fuel channel blockage accident considered as the Maximum Credible Accident for the reactor. The potential accident consequences and the criteria for radiological doses acceptance are evaluated and discussed. (author)

  15. HCTISN - Plenary extraordinary meeting on the 9 March 2012 - General consequences of the earthquake and tsunami; Status of Fukushima-Dai-ichi nuclear installations; The Fukushima accident, one year after: environmental and health situation in Japan; Protective actions undertaken by Japanese authorities; Support by AREVA to Japan after the Fukushima accident; What went on in Fukushima? Implementation of the IAEA nuclear safety action plan; Review of European stress tests by the peers; Opinion of the ASN on complementary safety assessments (CSAs); HCTISN - Reunion pleniere extraordinaire du 9 mars 2012: Consequences generales du seisme et du tusnami; Situation des installations nucleaires de Fukushima Dai-ichi; L'accident de Fukushima 1 an apres: situation environnementale et sanitaire au Japon; Les actions de protection engagees par les autorites japonaises; Aide apportee par AREVA au Japon suite a l'accident de Fukushima; Que s'est-il passe a Fukushima?

    Energy Technology Data Exchange (ETDEWEB)

    Kataoka, Susumu [Ambassade du Japon en France, 7 Avenue Hoche, 75008 Paris (France); Charles, T.; Champion, Didier [Institut de radioprotection et de surete nucleaire - IRSN, 31, avenue de la Division Leclerc, 92260 Fontenay-aux-Roses (France); Jean-Luc Godet [Autorite de surete nucleaire, 6, place du Colonel Bourgoin, 75012 Paris (France); ASN/DIS, 10, Route du Panorama, 92266 Fontenay-aux-Roses cedex (France); Arnaud GAY [Business Unit Valorisation - AREVA (France); Philippe Jamet [European Nuclear Safety Regulators Group - ENSREG, Autorite de surete nucleaire, 6, place du Colonel Bourgoin, 75012 Paris (France)

    2012-03-09

    This document contains Power Point presentations proposed during a plenary session of the High Committee transparency and information on nuclear safety (HCTISN). The contributions addressed the Fukushima accident (the earthquake and the tsunami, the technical consequences on the plant, the consequences on the environment and on health, the different actions undertaken in Japan to protect the population, the consequences on nuclear safety in other countries with notably the performance of stress tests or the organisation of complementary safety assessments on the French fleet of nuclear reactors

  16. [Emergency care for traffic accidents in Bavaria: current process analysis depending on hospital and emergency service structures].

    Science.gov (United States)

    Lackner, C K; Bielmeier, S; Burghofer, K

    2010-03-01

    A change is emerging in the hospital landscape due to health political measures, which in consequence also influences the prehospital medical care in emergencies. The main focus of this study was to gather information about emergency medical care after traffic accidents on the basis of data from Bavarian emergency medical services. In 2006 there were 14,261 traffic accidents in Bavaria where an emergency doctor attended the scene. The patients were primarily cared for by land-based rescue services and air rescue services were only used in 19.1% of the cases. Of the patients involved in a traffic accident 47.6% were transported to a primary health care hospital. A prehospital interval of more than 60 min occurred in 20% of the missions. Of the patients 96.2% were transported to tertiary or maximum care hospital by air rescue services but emergency facilities were, however restricted to daylight hours. There was a further limitation due to the routine duty hours in hospitals as only 36.7% of accidents occurred during this time intervall. An increase of admission post trauma in maximum care clinics occurred from 2002 until 2006 while simultaneously the prehospital period was extended. In order to assure sufficient trauma care for seriously injured persons a continuous 24 h availability of emergency trauma facilities is necessary. For this purpose it is necessary to establish regional trauma networks between receiving hospitals as well as air rescue services at night time. Furthermore, a cost-efficient compensation of the structural, personnel and logistic expenses for the treatment of the severely injured has to be assured.

  17. THE ANALYSIS OF AFFECT OF THE TRAFFIC ACCIDENTS ENDING DEATH TO THE LIFE EXPECTATIONS

    Directory of Open Access Journals (Sweden)

    Derya KOC

    2006-02-01

    Full Text Available We make observations for year 1999, backward on the traffic accidents and deaths in the registrations and countries, based on the age and sex classifications by analyzing using life table techniques. In conclusion we observed that in tables that are prepared by excluded traffic accidents, effects seen in all ranges of sex and totally life losing from the age of zero. Life expectancy for year 1999 in Turkey, our observations pointed out the number 74.43 on the life table including all the deaths, 74.61 on the life table excluding accidents. When we make observations on the sex stage, for men we finded out 72.00 on the life table including deaths 74.24 on the life table excluding accidents. For women we observed the numbers 76.99 and 77.10 on the life tables. [TAF Prev Med Bull 2006; 5(1.000: 32-40

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

    Science.gov (United States)

    Bilban, Marjan

    2002-12-01

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

  19. An epidemiologic survey of road traffic accidents in Iran: analysis of driver-related factors

    Directory of Open Access Journals (Sweden)

    Moafian Ghasem

    2013-06-01

    Full Text Available 【Abstract】Objective: Road traffic accident (RTA and its related injuries contribute to a significant portion of the burden of diseases in Iran. This paper explores the as-sociation between driver-related factors and RTA in the country. Methods: This cross-sectional study was conducted in Iran and all data regarding RTAs from March 20, 2010 to June 10, 2010 were obtained from the Traffic Police Department. We included 538 588 RTA records, which were classified to control for the main confounders: accident type, final cause of accident, time of accident and driver-related factors. Driver-related factors included sex, educational level, license type, type of injury, duration between accident and getting the driving license and driver’s error type. Results: A total of 538 588 drivers (91.83% male, sex ratio of almost 13:1 were involved in the RTAs. Among them 423 932 (78.71% were uninjured; 224 818 (41.74% had a diploma degree. Grade 2 driving license represented the highest proportion of all driving licenses (290 811, 54.00%. The greatest number of accidents took place at 12:00-13:59 (75 024, 13.93%. The proportion of drivers involved in RTAs decreased from 15.90% in the first year of getting a driving license to 3.13% after 10 years’ of driving experience. Ne-glect of regulations was the commonest cause of traffic crashes (345 589, 64.17%. Non-observance of priority and inattention to the front were the most frequent final causes of death (138 175, 25.66% and 129 352, 24.02%, respectively. We found significant association between type of acci-dent and sex, education, license type, time of accident, final cause of accident, driver’s error as well as duration between accident and getting the driving license (all P<0.001. Conclusion: Our results will improve the traffic law enforcement measures, which will change inappropriate be-havior of drivers and protect the least experienced road users. Key words: Accidents, traffic; Automobile

  20. GIS based analysis of Intercity Fatal Road Traffic Accidents in Iran.

    Science.gov (United States)

    Alizadeh, A; Zare, M; Darparesh, M; Mohseni, S; Soleimani-Ahmadi, M

    2015-01-01

    Road traffic accidents including intercity car traffic accidents (ICTAs) are among the most important causes of morbidity and mortality due to the growing number of vehicles, risky behaviors, and changes in lifestyle of the general population. A sound knowledge of the geographical distribution of car traffic accidents can be considered as an approach towards the accident causation and it can be used as an administrative tool in allocating the sources for traffic accidents prevention. This study was conducted to investigate the geographical distribution and the time trend of fatal intercity car traffic accidents in Iran. To conduct this descriptive study, all Iranian intercity road traffic mortality data were obtained from the Police reports in the Statistical Yearbook of the Governor's Budget and Planning. The obtained data were for 17 complete Iranian calendar years from March 1997 to March 2012. The incidence rate (IR) of fatal ICTAs for each year was calculated as the total number of fatal ICTAs in every 100000 population in specified time intervals. Figures and maps indicating the trends and geographical distribution of fatal ICTAs were prepared while using Microsoft Excel and ArcGis9.2 software. The number of fatal car accidents showed a general increasing trend from 3000 in 1996 to 13500 in 2012. The incidence of fatal intercity car accidents has changed from six in 100000 population in 1996 to 18 in 100000 population in 2012. GIS based data showed that the incidence rate of ICTAs in different provinces of Iran was very divergent. The highest incidence of fatal ICTAs was in Semnan province (IR= 35.2), followed by North Khorasan (IR=22.7), and South Khorasan (IR=22). The least incidence of fatal ICTAs was in Tehran province (IR=2.4) followed by Khozestan (IR=6.5), and Eastern Azarbayejan (IR=6.6). The compensation cost of fatal ICTAs also showed an increasing trend during the studied period. Since an increasing amount of money was being paid yearly for the car

  1. 日本福岛第一核电站事故源项及后果评价%Fukushima Daiichi NPS Accident Source Term and Consequence Assessment

    Institute of Scientific and Technical Information of China (English)

    王海洋; 黄树明; 王晓霞; 尤伟; 米爱军; 张普忠

    2011-01-01

    根据已有的日本福岛第一核电站相关资料,利用美国核管理委员会《轻水堆核电厂事故源项》中的假设条件,计算出事故后安全壳内的放射性源项,综合考虑各种不确定性因素,得出较为保守的环境释放源项。采用美国核管理委员会RG 1.4中大气扩散模式的假设计算大气弥散因子,并应用ICRP 71号出版物F、GR 12号报告等资料中的剂量计算模式及剂量转换因子进行了事故剂量后果的估算、分析和评价。%Conservative source term of radioactive release to the environment is calculated after the accident at Fukushima Daiichi nuclear power station based on updated informational materials and the assumptions made in NUREG-1465 by U.S.Nuclear Regulatory Commission(U.S.NRC),with the consideration of the various uncertainties.And a series of atmospheric dispersion factors are obtained from U.S.NRC Regulatory Guide 1.4(RG 1.4).Finally,this paper provides calculation of the accidental dose,which is analyzed and assessed specifically,using the models and parameters in ICRP Publication 71,FGR 12 and so on.

  2. Dermatological consequences of the Cs-137 radiological accident in Goiania, Goias State, Brazil; Repercussoes dermatologicas no acidente radioativo com o Cesio 137 em Goiania

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Lia Candida Miranda de

    1996-07-01

    The objective of the present study was to analyse the occurrence of dermatosis in individuals that had been exposed to cesium{sup 137} during the radioactive accident in Goiania, in 1987 and detect pre-cancerous dermatosis or those predictive of low immunity. The groups were evaluated according to the intensity of radiation they had been exposed to and then compared to a control group of people not exposed to radiation. The population exposed to the cesium{sup 137} was comprised of 109 people, who were divided into Groups I and II, according to the CNEN norms. In group I, 54 people with {<=} 20 rads exposure and/or radio lesion were included; in group II, 55 people with > 20 rads exposure were included, along with the children of group I individuals. This was a historic cohort study, that is, a retrospective study that lasted 9 years, extending from September of 1987 to August, 1996. The presence of the oncoprotein p-53 was studied in the radio lesions of 10 patients. There is no evidence of an increase in the incidence of dermatosis in the exposed groups, excepts for pyoderma in patients with radio lesions. The most frequent dermatosis were: pyoderma, pityriasis versicolor, scabies, dermatophytosis and seborrhoeic dermatitis. The results obtained were not statistically significant for the evaluation of dermatosis predictive of low immunity or precancerous lesions. The oncoprotein p-53 in individuals with radio lesion showed a 80% positivity rate and risk factor estimated in 8 times, for the test. It has proved to be useful because it represents one more option in terms of propaedeutic evaluation and suggests that one should pay close and continuous attention in order to better control the evolution of these individuals. (author)

  3. Measuring Risk Aversion for Nuclear Power Plant Accident: Results of Contingent Valuation Survey in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of nuclear power plant (NPP) severe accident is one of the major topics to be addressed. For the evaluation of the external cost of NPP severe accident, the effect of public risk averse behavior against the group accidents, such as NPP accident, dam failure, must be addressed. Although the equivalent fatalities from a single group accident are not common and its risk is very small compared to other accidents, people perceive the group accident more seriously. In other words, people are more concerned about low probability/high consequence events than about high probability/low consequence events having the same mean damage. One of the representative method to integrate the risk aversion in the external costs of severe nuclear reactor accidents was developed by Eeckoudt et al., and he used the risk aversion coefficient, mainly based on the analysis of financial risks in the stock markets to evaluate the external cost of nuclear severe accident. However, the use of financial risk aversion coefficient to nuclear severe accidents is not appropriate, because financial risk and nuclear severe accident risk are entirely different. In this paper, the individual-level survey was conducted to measure the risk aversion coefficient and estimate the multiplication factor to integrate the risk aversion in the external costs of NPP severe accident. This study propose an integrated framework on estimation of the external cost associated with severe accidents of NPP considering public risk aversion behavior. The theoretical framework to estimate the risk aversion coefficient/multiplication factor and to assess economic damages from a hypothetical NPP accident was constructed. Based on the theoretical framework, the risk aversion coefficient can be analyzed by conducting public survey with a carefully designed lottery questions. Compared to the previous studies on estimation of the

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

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, N.

    2011-02-14

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

  5. Development of Auditing Technology for Accident Analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, Y. J.; Jeong, J. J.; Kim, H. C.; Chung, Y. J.; Bae, K. H

    2006-02-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. At initial period, PIRT has been performed to identify the model deficiencies and determine the priority of model improvements. The identified thermal hydraulic models has been implemented to RELAP5/MOD3.3 auditing code according to the PIRT ranking. The input model for SMART-P has been developed with consistent to the current design status documents and checked by independent reviewer as Q/A procedure.The evaluation of experimental availabilities and code collapsible has been done by expert group and summarized as validation matrix forms. The experimental data of VISTA, which is the only integral effect test facility, were used to validate the improved model. The safety analysis has been demonstrated for the essential accident scenario. The validation and demonstration show that the developed models are applicable to utilize in reliable and independent auditing for SMART design certification.

  6. Revealing the association between cerebrovascular accidents and ambient temperature: a meta-analysis

    Science.gov (United States)

    Zorrilla-Vaca, Andrés; Healy, Ryan Jacob; Silva-Medina, Melissa M.

    2016-10-01

    The association between cerebrovascular accidents (CVA) and weather has been described across several studies showing multiple conflicting results. In this paper, we aim to conduct a meta-analysis to further clarify this association, as well as to find the potential sources of heterogeneity. PubMed, EMBASE, and Google Scholar were searched from inception through 2015, for articles analyzing the correlation between the incidence of CVA and temperature. A pooled effect size (ES) was estimated using random effects model and expressed as absolute values. Subgroup analyses by type of CVA were also performed. Heterogeneity and influence of covariates—including geographic latitude of the study site, male percentage, average temperature, and time interval—were assessed by meta-regression analysis. Twenty-six articles underwent full data extraction and scoring. A total of 19,736 subjects with CVA from 12 different countries were included and grouped as ischemic strokes (IS; n = 14,199), intracerebral hemorrhages (ICH; n = 3798), and subarachnoid hemorrhages (SAH; n = 1739). Lower ambient temperature was significantly associated with increase in incidence of overall CVA when using unadjusted (pooled ES = 0.23, P < 0.001) and adjusted data (pooled ES = 0.03, P = 0.003). Subgroup analyses showed that lower temperature has higher impact on the incidence of ICH (pooled ES = 0.34, P < 0.001), than that of IS (pooled ES = 0.22, P < 0.001) and SAH (pooled ES = 0.11, P = 0.012). In meta-regression analysis, the geographic latitude of the study site was the most influencing factor on this association (Z-score = 8.68). Synthesis of the existing data provides evidence supporting that a lower ambient temperature increases the incidence of CVA. Further population-based studies conducted at negative latitudes are needed to clarify the influence of this factor.

  7. An epidemiologic survey of road traffic accidents in Iran:analysis of driver-related factors

    Institute of Scientific and Technical Information of China (English)

    Ghasem Moafian; Mohammad Reza Aghabeigi; Seyed Taghi Heydari; Amin Hoseinzadeh; Kamran Bagheri Lankarani; Yaser Sarikhani

    2013-01-01

    Road traffic accident (RTA)and its related injuries contribute to a significant portion of the burden of diseases in Iran.This paper explores the association between driver-related factors and RTA in the country.Methods:This cross-sectional study was conducted in Iran and all data regarding RTAs from March 20,2010 to June 10,2010 were obtained from the Traffic Police Department.We included 538 588 RTA records,which were classified to control for the main confounders:accident type,final cause of accident,time of accident and driver-related factors.Driver-related factors included sex,educational level,license type,type of injury,duration between accident and getting the driving license and driver's error type.Results:Atotal of 538 588 drivers (91.83% male,sex ratio of almost 13:1) were involved in the RTAs.Among them 423 932 (78.71%) were uninjured; 224 818 (41.74%) had a diploma degree.Grade 2 driving license represented the highest proportion of all driving licenses (290 811,54.00%).The greatest number of accidents took place at 12:00-13:59(75 024,13.93%).The proportion of drivers involved in RTAs decreased from 15.90% in the first year of getting a driving license to 3.13% after 10 years' of driving experience.Neglect of regulations was the commonest cause of traffic crashes (345 589,64.17%).Non-observance of priority and inattention to the front were the most frequent final causes of death (138 175,25.66% and 129 352,24.02%,respectively).We found significant association between type of accident and sex,education,license type,time of accident,final cause of accident,driver's error as well as duration between accident and getting the driving license (all P<0.001).Conclusion:Our results will improve the traffic law enforcement measures,which will change inappropriate behavior of drivers and protect the least experienced road users.

  8. SCAP: a new methodology for safety management based on feedback from credible accident-probabilistic fault tree analysis system.

    Science.gov (United States)

    Khan, F I; Iqbal, A; Ramesh, N; Abbasi, S A

    2001-10-12

    As it is conventionally done, strategies for incorporating accident--prevention measures in any hazardous chemical process industry are developed on the basis of input from risk assessment. However, the two steps-- risk assessment and hazard reduction (or safety) measures--are not linked interactively in the existing methodologies. This prevents a quantitative assessment of the impacts of safety measures on risk control. We have made an attempt to develop a methodology in which risk assessment steps are interactively linked with implementation of safety measures. The resultant system tells us the extent of reduction of risk by each successive safety measure. It also tells based on sophisticated maximum credible accident analysis (MCAA) and probabilistic fault tree analysis (PFTA) whether a given unit can ever be made 'safe'. The application of the methodology has been illustrated with a case study.

  9. Determinants of participation in a longitudinal two-stage study of the health consequences of the Chornobyl nuclear power plant accident

    Directory of Open Access Journals (Sweden)

    Zakhozha Victoria

    2008-05-01

    Full Text Available Abstract Background The determinants of participation in long-term follow-up studies of disasters have rarely been delineated. Even less is known from studies of events that occurred in eastern Europe. We examined the factors associated with participation in a longitudinal two-stage study conducted in Kyiv following the 1986 Chornobyl nuclear power plant accident. Methods Six hundred child-mother dyads (300 evacuees and 300 classmate controls were initially assessed in 1997 when the children were 11 years old, and followed up in 2005–6 when they were 19 years old. A population control group (304 mothers and 327 children was added in 2005–6. Each assessment point involved home interviews with the children and mothers (stage 1, followed by medical examinations of the children at a clinic (stage 2. Background characteristics, health status, and Chornobyl risk perceptions were examined. Results The participation rates in the follow-up home interviews were 87.8% for the children (88.6% for evacuees; 87.0% for classmates and 83.7% for their mothers (86.4% for evacuees and 81.0% for classmates. Children's and mothers' participation was predicted by one another's study participation and attendance at the medical examination at time 1. Mother's participation was also predicted by initial concerns about her child's health, greater psychological distress, and Chornobyl risk perceptions. In 1997, 91.2% of the children had a medical examination (91.7% of evacuees and 90.7% of classmates; in 2005–6, 85.2% were examined (83.0% of evacuees, 87.7% of classmates, 85.0% of population controls. At both times, poor health perceptions were associated with receiving a medical examination. In 2005–6, clinic attendance was also associated with the young adults' risk perceptions, depression or generalized anxiety disorder, lower standard of living, and female gender. Conclusion Despite our low attrition rates, we identified several determinants of selective

  10. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  11. 一起氧气串入氮气管网的事故分析%A string of oxygen to the nitrogen gas pipe network of accident analysis

    Institute of Scientific and Technical Information of China (English)

    安浩龙

    2015-01-01

    针对外压缩流程20000m3/h氧压机出口氧气串入氮气管网事故进行回顾,根据现场工艺管道的布置及后系统氮气使用情况,分析氧气串入氮气管网事故发生的原因及有可能造成的后果,并提出相应的预防措施.%For the compression process after 20000 m3/h oxygen compressor series with oxygen nitrogen export pipeline accidents are reviewed, according to the process piping layout and system nitrogen usage, the analysis of the series with oxygen nitrogen pipeline accident causes and possible consequences,and the corresponding preventive measures are proposed.

  12. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  15. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

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

    Energy Technology Data Exchange (ETDEWEB)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

    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.

  17. Safety Assessment of High-Risk Operations in Hydroelectric-Project Based on Accidents Analysis, SEM, and ANP

    Directory of Open Access Journals (Sweden)

    Jian-Lan Zhou

    2013-01-01

    Full Text Available Safety risk analysis and assessment of high-risk work system in hydroelectric project has an important role in safety management. The interactive relationships between human factors and the importance of factors are analyzed and proposed. We analyze the correlation relationship among the factors by using statistical method, which is more objective than subjective judgment. The HFACS is provided to establish a rational and an applicable index system for investigating human error in accidents; the structural equation modeling (SEM and accident data are used to construct system model and acquire the path coefficient among the risk factor variables; the ANP model is built to assess the importance of accident factors. 289 pieces of valid questionnaires data are analyzed to obtain the path coefficient between risk factor variables and to build the ANP model’s judgment matrix. Finally, the human factors’ weights are calculated by ANP model. Combining SEM’s results and factor's frequency analysis and building the ANP model, the results show that the four greatest weight values of the factors are, respectively, “personal readiness,” “perception and decision errors,” “skill-based errors,” and “violation operations.” The results of ANP model provide a reference for the engineering and construction management.

  18. Industrial accidents triggered by lightning.

    Science.gov (United States)

    Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

    2010-12-15

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

  19. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Science.gov (United States)

    2010-07-01

    ... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.8 Access to off-site consequence analysis information by Federal... analysis information by Federal government officials. 1400.8 Section 1400.8 Protection of...

  20. Risk Analysis for Public Consumption: Media Coverage of the Ginna Nuclear Reactor Accident.

    Science.gov (United States)

    Dunwoody, Sharon; And Others

    Researchers have determined that the lay public makes risk judgments in ways that are very different from those advocated by scientists. Noting that these differences have caused considerable concern among those who promote and regulate health and safety, a study examined media coverage of the accident at the Robert E. Ginna nuclear power plant…

  1. An Analysis of Station Blackout Sequences Using MELCOR1.8.5 Code for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y. M.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    The Korea Atomic Energy Research Institute (KAERI) has been constructing severe accident analysis database (DB) under a National Nuclear R and D Program. Especially, MAAP (commercial code being widely used for industries) DB for many scenarios including station blackout (SBO) has been completed up to now. This report shows the analysis results for SBO scenarios using MELCOR code. These results will be used for the degree of completion after being compared with MAAP results. The developing strategy of MELCOR code is the same with that of MAAP DB. For the generation of data set, the Korean standard nuclear power plant (KSNP) has been selected as a reference plant and the eight SBO scenarios are chosen to be analyzed based on the PSA results (these eight scenarios accounted for 99 percent of occurrence frequency of total 197 SBO scenarios). Both thermal hydraulics (T/H) and source term analysis have been performed using MELCOR version 1.8.5 for the chosen scenarios. But only major T/H variables treated in the MAAP report are listed among the generated data set, which shows the characteristics of each scenario. These SBO results together with those of the other initiating events (to be analyzed in the future) will be used as inputs for DB construction and special value will be found in the comparing and complimentary process with MAAP DB

  2. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  3. A Longitudinal Analysis of the Causal Factors in Major Maritime Accidents in the USA and Canada (1996-2006)

    Science.gov (United States)

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

    2007-01-01

    Accident reports provide important insights into the causes and contributory factors leading to particular adverse events. In contrast, this paper provides an analysis that extends across the findings presented over ten years investigations into maritime accidents by both the US National Transportation Safety Board (NTSB) and Canadian Transportation Safety Board (TSB). The purpose of the study was to assess the comparative frequency of a range of causal factors in the reporting of adverse events. In order to communicate our findings, we introduce J-H graphs as a means of representing the proportion of causes and contributory factors associated with human error, equipment failure and other high level classifications in longitudinal studies of accident reports. Our results suggest the proportion of causal and contributory factors attributable to direct human error may be very much smaller than has been suggested elsewhere in the human factors literature. In contrast, more attention should be paid to wider systemic issues, including the managerial and regulatory context of maritime operations.

  4. Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Amos, C.N.; Wagner, K.C.; Bradley, D.R. (Science Applications International Corp., Albuquerque, NM (United States))

    1992-01-01

    A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained.

  5. Sensitivity Analysis of Dousing Spray Trip on Radioactive Release in Pressure Tube Rupture Accident with Both End Fitting Failures

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M. S.; Kang, H. S; Kim, S. R. [NESS, Daejeon (Korea, Republic of)

    2015-10-15

    We analyzed the sensitivity analysis of dousing spray trip conditions on radioactive release. In terms of conservativeness, the set 1 trip would be more appropriate in RR analysis than set 2 trip, which is the general condition of RR analysis. Radioactive releases from the containment building is related to containment air pressure, which increases by the coolant discharge from loss of coolant accident and the actuation conditions of dousing spray and so on. In LOCA analysis, the dousing spray trip conditions are set for the analysis objectives; for peak pressure (PP), for pressure signal (PS), for radioactive release (RR) and etc. In RR analysis, we would determine the dousing spray trip condition to increase radioactive release to the public for conservatism. Therefore, we carried out the sensitivity analysis of dousing spray trip condition on radioactive release from containment building using GOTHIC and SMART program for CANDU.

  6. Time analysis of fatal traffic accidents in Fars Province of Iran

    Directory of Open Access Journals (Sweden)

    Heydari Seyed Taghi

    2013-04-01

    Full Text Available 【Abstract】 Objective: To analyze the time factor in road traffic accidents (RTAs in Fars Province of Iran. Methods: This study was conducted in Fars Province, Iran from November 22, 2009 to November 21, 2011. Victims’ information consisted of age, sex, death toll involving dri-vers or passengers of cars, motorcycles and pedestrians, and site of injury etc. Accidents were analyzed in relation to hour of the day, season of the year, lighting condition in-cluding sunrise, sunset, daytime and nighttime. Results: A total of 3 642 deaths (78.3% were males, and the ratio of males to females was about 3.6:1 were studied regarding their autopsy records. There was a steady in-crease in fatal accidents occurring at midnight to 15:59. The risk of being involved in a fatal traffic accident was higher for those injured between 4:00 to 7:59 than at other times (OR=2.13, 95% CI 1.85-2.44. The greatest number of fatal RTAs took place in summer. Mortalities due to RTA during spring and summer were more pronounced at 20:00 to 23:59 and midnight to 3:59, whereas mortalities in fall and winter were more pronounced from 12:00 to 15:59. Conclusion: The high mortality rate of RTA is a major public health problem in Fars Province. Our results indicate that the time is an important factor which contributes to road traffic deaths. Key words: Accidents, traffic; Epidemiology; Mortality; Iran

  7. Retrospective reconstruction of Ioidne-131 distribution at the Fukushima Daiichi Nuclear Power Plant accident by analysis of Ioidne-129

    Science.gov (United States)

    Matsuzaki, Hiroyuki; Muramatsu, Yasuyuki; Toyama, Chiaki; Ohno, Takeshi; Kusuno, Haruka; Miyake, Yasuto; Honda, Maki

    2014-05-01

    Science and Education on June, 2011. So far more than 500 samples were measured and determined I-129 deposition amount by AMS at MALT (Micro Analysis Laboratory, Tandem accelerator), The University of Tokyo. The measurement error from AMS is less than 5%, typically 3%. The overall uncertainty is estimated less than 30%, including the uncertainty from that of the nominal value of the standard reference material used, that of I-129/I-131 ratio estimation, that of the "representativeness" for the region by the analyzed sample, etc. The isotopic ratio I-129/I-131 from the reactor was estimated [3] (to be 22.3 +- 6.3 as of March 11, 2011) from a series of samples collected by a group of The University of Tokyo on the 20th of April, 2011 for which the I-131 was determined by gamma-ray spectrometry with good precision. Complementarily, we had investigated the depth profile in soil of the accident derived I-129 and migration speed after the deposition and found that more than 90% of I-129 was concentrated within top 5 cm layer and the downward migration speed was less than 1cm/yr [4]. From the set of I-129 data, corresponding I-131 were calculated and the distribution map is going to be constructed. Various fine structures of the distribution came in sight. [1] Y. Nikiforov and D. R. Gnepp, 1994, Cancer, Vol. 47, pp748-766. [2] T. Straume, et al., 1996, Health Physics, Vol. 71, pp733-740. [3] Y. Miyake, H. Matsuzaki et al.,2012, Geochem. J., Vol. 46, pp327-333. [4] M. Honda, H. Matsuzaki et al., under submission.

  8. Environmental Aftermath of the Radiation Accident at Tomsk-7

    Science.gov (United States)

    Porfiriev, Boris N.; Porfiriev, Boris N.

    1996-01-01

    An analysis is presented of the environmental effects of the most serious radiation accident recorded after Chernobyl, which occurred in the formerly secret town of Tomsk-7 in Siberia, Russia, on 6, April 1993. Fortunately, it appears not to have become a major industrial crisis or disaster. The causes of the accident are described. It is argued that a mixture of both objective and subjective prerequisites, including specific human, organizational, and technological factors, were responsible for the explosion or directly facilitated it. The Tomsk-7 accident’s ecological, medical, social, and psychological consequences are discussed.

  9. Effects of source term characteristics on Off Site consequence

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Off site consequence analysis in Level 3 PSA is mainly affected by source term release characteristics of nuclear plant. The severe accident analysis codes for quantifying the source term release characteristics such as MELCOR and MAAP provide detailed information of these characteristics to assess the off site consequence. The aforementioned characteristics, however, have not been considered in the consequence analysis of domestic plants because of large uncertainty in these characteristics so far. Recently, the USNRC SOARCA report showed an approach to utilize detailed source term characteristics provided by MELCOR code to quantify the off site consequence more realistically. Main purpose of this study is to assess effects of the MELCOR source term characteristics on the off site consequence analysis of a domestic nuclear power plant, in a similar fashion to the SOARCA approach. Among many features characterizing source term, the most important one is to determine initial and boundary conditions of atmospheric dispersion such as:- Release amounts of source term - Release time and duration Moreover, plumes features (i.e., radiation clouds) affect atmospheric dispersion that shapes plume characteristics as follows: - Initial dimension of plumes - Plume rise - Deposition of radioactive materials during dispersion Although the current severe accident codes have some limitation in providing the entire source term release characteristics needed in the consequence analysis, the essential information for these features could be obtained from these codes. It is noted that the typical approaches, which generate source term information for the consequence analysis from the severe accident codes, should require a technical manipulation by the experts of consequence analysis. The present effort focused on an identification of insights to utilize source term characteristics of the severe accident codes.

  10. Steady state and accident analysis of SCOR (simple compact reactor) with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Marie-Sophie Chenaud; Guy-Marie Gautier [CEA Cadarache- 13108 St Paul Lez Durance (France)

    2005-07-01

    Full text of publication follows: Within the framework of innovative reactors studies, the CEA was led to propose the SCOR design (Simple Compact Reactor). This design is based on a compact 600 MWe PWR and combines most of the advantages of innovative reactors. All main components such as the pressurizer, the canned pumps, the control rod mechanics and the dedicated heat exchangers on the passive residual heat removal system are integrated in the vessel.The only steam generator is located above the vessel in place of the upper head. The reactor operates at much lower primary circuit pressure than standard PWRs (85 bar instead of the usual 155 bar) and the power density is low (70 MW/m{sup 3} instead of 100 MW/m{sup 3} for the present PWRs). The reactivity being controlled by control rods and burnable poisons, there is no soluble boron. The elimination of a serious LOCA (Loss Of Coolant Accident) and the integrated residual heat removal system lead to enhanced safety with simple safety systems. Main features of the SCOR design and functional parameters have been previously reported. This paper focuses on the safety analysis of SCOR. Thermo hydraulic calculations have been run with the CATHARE code. Some calculations were run with the point kinetics module of CATHARE. Several transient simulations have been assessed. They concern a normal reactor trip from full power operation till refueling shutdown and accidental scenarios such as: - Loss of power, - Breaks from 0.02 m to 0.1 m on circuits connected to the vessel, - Steam generator tubes rupture, - Reactivity insertion by cold shock. Results of transient simulations enable us to conclude upon: - the increase of grace periods in comparison with standard PWRs if no safety systems operate besides emergency shutdown, - the expected efficiency of designed safety systems and in particular of the residual heat removal system in passive configuration even when integrated exchanger are dewatered. It will be retained that

  11. The path of accident analysis: the traditional paradigm and extending the origins of the expansion of analysis

    Directory of Open Access Journals (Sweden)

    Ildeberto Muniz de Almeida

    2006-01-01

    Full Text Available The traditional approach to accidents assumes that compliance with procedures and norms protects the system from accidents and that these events are caused by the faulty behavior of workers, which results partly from personality aspects. Identification of these behaviors can be based on comparing them with the standard "safe working practices", which safety experts are aware of ahead of time. In recent decades, new alternative views have expanded the perimeters of accident analyses and opened the way to questioning the assumption of the traditional approach to the concepts of the human being and work. These new approaches help to highlight the sterile results of traditional practices: blaming and punishing victims, recommending training, and proposing norms without changing the systems in which the accidents took place. The new approaches suggest that the traditional approach is totally worn out and emphasize the importance of operator contribution for system safety.

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

    CERN Document Server

    Balda, F

    2002-01-01

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

  13. Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

    OpenAIRE

    Kil-Mo Koo; Jin-Ho Song; Sang-Baik Kim; Kwang-Il Ahn; Won-Pil Baek; Kil-Nam Oh; Gyu-Tae Kim

    2012-01-01

    Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity) occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard...

  14. Time analysis of fatal traffic accidents in Fars Province of Iran

    Institute of Scientific and Technical Information of China (English)

    Seyed Taghi Heydari; Amin Hoseinzadeh; Yaser Sarikhani; Arya Hedjazi; Mohammad Zarenezhad; Ghasem Moafian; Mohammad Reza Aghabeigi

    2013-01-01

    Objective:To analyze the time factor in road traffic accidents (RTAs) in Fars Province of Iran.Methods:This study was conducted in Fars Province,Iran from November 22,2009 to November 21,2011.Victims'information consisted of age,sex,death toll involving drivers or passengers of cars,motorcycles and pedestrians,and site of injury etc.Accidents were analyzed in relation to hour of the day,season of the year,lighting condition including sunrise,sunset,daytime and nighttime.Results:Atotal of 3 642 deaths (78.3% were males,and the ratio of males to females was about 3.6:1) were studied regarding their autopsy records.There was a steady increase in fatal accidents occurring at midnight to 15:59.The risk of being involved in a fatal traffic accident was higher for those injured between 4:00 to 7:59 than at other times (OR=2.13,95% CI 1.85-2.44).The greatest number of fatal RTAs took place in summer.Mortalities due to RTA during spring and summer were more pronounced at 20:00 to 23:59and midnight to 3:59,whereas mortalities in fall and winter were more pronounced from 12:00 to 15:59.Conclusion:The high mortality rate ofRTAis a major public health problem in Fars Province.Our results indicate that the time is an important factor which contributes to road traffic deaths.

  15. Epidemiological analysis of traffic accident trauma in Gansu province in 1996

    Institute of Scientific and Technical Information of China (English)

    张向东; 代荫梅

    1999-01-01

    Objective To analyze the epidemiologic data of traffic trauma in Gansu province in 1996 and try to find effective ways to reduce the injury.Methods The data were gathered from the General Team o the Traffic Police of Gansu Province and analyzed together with other related data.Results Although the traffic accidents in Gansu province were reduced in last two years as a result of traffic safety education,the number of casuatlties has not been evidently reduced.The number of deaths caused by traffic accidents was 983 in 1996.The main causes of these deaths were the negligence of the drivers,carelessness of the pedestrians and the bike riders,the sudden breakdown of the machine parts of the vehicles,and non-licensed driving.Among the number of deaths 69 percent was caused by violation of traffic regulations by drivers.Most of the death accidents happened at straight roads and road-crosses.The percentage was 64% and 11%,respectively.The most of deaths,about 81%,took place in sunny days.The main reason was due to the careless and exceeding-speed-limit driving.The young and middle aged were about 77% of the dead,most likely because they are the dominant group in daily work and life.Conclusions To strengthen the propaganda of traffic regulations,improve driver's moral qualities and raise the management level are very important for reducing traffic accidents.Correct and timely first aid before being hospitalized can also greatly reduce the mortality.

  16. 3D analysis of the reactivity insertion accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  17. Risk Estimation Methodology for Launch Accidents.

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

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

    2014-03-01

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

  19. SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Brad J. Merrill; Shannon M Bragg-Sitton

    2013-09-01

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

  20. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  1. Dose rate mapping and quantitative analysis of radioactive deposition with simple monitoring instruments in Finland after the Chernobyl accident.

    Energy Technology Data Exchange (ETDEWEB)

    Koivukoski, J. [Ministry of the Interior, Rescue Dept., Helsinki (Finland); Paatero, J. [Finnish Meteorological Inst., Helsinki (Finland)], E-mail: janne.koivukoski@intermin.fi

    2013-03-01

    This article reviews the Finnish dose-rate mapping equipment and the system to process the obtained results, which were used immediately after the 1986 Chernobyl accident. We present the results of the external gamma-radiation monitoring carried out with simple civil-defence gamma monitoring instruments and compare them with the subsequent deposition mapping performed with research-grade instruments. The analysis shows that the quality of radiation mapping is good enough for decision makers to direct protective measures to the right areas. This review also demonstrates that a simple stationary external gamma radiation monitoring network can be effectively used for early warning in radiation emergency situations. (orig.)

  2. An analysis of thermionic space nuclear reactor power system: I. Effect of disassembling radial reflector, following a reactivity initiated accident

    Science.gov (United States)

    El-Genk, Mohamed S.; Paramonov, Dmitry

    1993-01-01

    An analysis is performed to determine the effect of disassembling the radial reflector of the TOPAZ-II reactor, following a hypothetical severe Reactivity Initiated Accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the drums to rotate the full 180° outward at their maximum speed of 1.4°/s. Results indicate that disassembling only three of twelve radial reflector panels would successfully shutdown the reactor, with little overheating of the fuel and the moderator.

  3. Probabilistic consequence analysis for vapor cloud explosion of flammable gas%可燃气云爆炸事故后果的概率分析

    Institute of Scientific and Technical Information of China (English)

    陈国华; 周剑峰; 张文海; 张晖; 陈清光

    2006-01-01

    Consequence analysis is very important for risk evaluation. Vapor cloud explosion (VCE) is one of the main accident types of flammable gas. Because of the limitation of people's knowledge about accident consequence, influence of natural environment, and complex process of the development of an accident, some parameters of the VCE models can not be precisely defined, so that different input values can make great difference in the final results. A probabilistic consequence analysis method based on Monte-Carlo simulation (MCS) is proposed to help analyze the influence of an accident. How to evaluate the fatality radius when its probability is given is also analyzed, a regression estimate method based on Support Vector Machine (SVM) is proposed to solve this problem. The consequence severity can be reflected by the extent of damage (fatality radius, FR) and its related probability. Input arguments which have stochastic characteristic and their probability distributions are analyzed, and the application of the probabilistic consequence analysis method for VCE of LPG (liquefied petroleum gas) is illustrated.%对事故后果进行分析,是安全评价的重要内容.可燃气体泄漏后发生蒸气云爆炸是一种重要的事故类型,由于人们对事故后果认识的局限、环境的影响、以及事故发展过程的复杂性,VCE后果分析中许多参数未能准确定义,部份输入参数取值具有随机性,不同取值使最终的分析结果相差很大.本文提出了一种事故后果的概率分析方法,通过事故的后果(死亡半径)及其发生的概率来反映后果严重程度,并基于蒙特卡罗模拟方法来计算事故后果的发生概率.本文对给定发生概率时如何确定死亡半径也进行了分析,采用基于支持向量机(SVM)的非线性回归估计方法对死亡半径进行估计.分析了VCE后果分析模型中的随机性参数及其取值,结合具体的应用实例,说明了该方法在LPG储罐蒸气云爆炸后果分析中的应用.

  4. Consequences analysis of accidents in an air-propane plant; Analise de consequencias de acidentes numa planta de ar propanado

    Energy Technology Data Exchange (ETDEWEB)

    Lucena, Sergio; Lima Flho, Nelson M. de; Martins, Andrea do N. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Campos, Michel F. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil); Zimmerle, Sergio Ricardo T.S. [Companhia Pernambucana de Gas (COPERGAS), Recife, PE (Brazil); Alencar, Joao Rui B. de [LAFEPE - Laboratorio Farmaceutico do Estado de Pernambuco S/A, Recife, PE (Brazil); Martins, Andrea do N. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2004-07-01

    The Universidade Federal do Rio Grande do Norte has been developing prototype ovens, which work with natural gas. All the project of the prototypes, which will be applied to ceramic, bread bake industry and incineration of the hospital garbage, needs to be studied, developed and tested carefully until its conclusion. Then VRML language (Virtual Reality Modelling Language) is used as a tool in the study of the engineering projects and simulation of some tests. The main benefits of the use of this tool are: finding and solving problems in the project of the prototypes faster; optimization in the project since the three-dimensional visualization facilitates the study ; and simulation of aspects of functioning of the ovens before its construction. (author)

  5. Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

  6. Dementia and Traffic Accidents

    DEFF Research Database (Denmark)

    Petersen, Jindong Ding; Siersma, Volkert; Nielsen, Connie Thurøe;

    2016-01-01

    BACKGROUND: As a consequence of a rapid growth of an ageing population, more people with dementia are expected on the roads. Little is known about whether these people are at increased risk of road traffic-related accidents. OBJECTIVE: Our study aims to investigate the risk of road traffic......-related accidents for people aged 65 years or older with a diagnosis of dementia in Denmark. METHODS: We will conduct a nationwide population-based cohort study consisting of Danish people aged 65 or older living in Denmark as of January 1, 2008. The cohort is followed for 7 years (2008-2014). Individual's personal...... data are available in Danish registers and can be linked using a unique personal identification number. A person is identified with dementia if the person meets at least one of the following criteria: (1) a diagnosis of the disease in the Danish National Patient Register or in the Danish Psychiatric...

  7. Analysis of human error in occupational accidents in the power plant industries using combining innovative FTA and meta-heuristic algorithms

    Directory of Open Access Journals (Sweden)

    M. Omidvari

    2015-09-01

    Full Text Available Introduction: Occupational accidents are of the main issues in industries. It is necessary to identify the main root causes of accidents for their control. Several models have been proposed for determining the accidents root causes. FTA is one of the most widely used models which could graphically establish the root causes of accidents. The non-linear function is one of the main challenges in FTA compliance and in order to obtain the exact number, the meta-heuristic algorithms can be used. Material and Method: The present research was done in power plant industries in construction phase. In this study, a pattern for the analysis of human error in work-related accidents was provided by combination of neural network algorithms and FTA analytical model. Finally, using this pattern, the potential rate of all causes was determined. Result: The results showed that training, age, and non-compliance with safety principals in the workplace were the most important factors influencing human error in the occupational accident. Conclusion: According to the obtained results, it can be concluded that human errors can be greatly reduced by training, right choice of workers with regard to the type of occupations, and provision of appropriate safety conditions in the work place.

  8. SCENARIO OF AN ACCIDENT OF SOIL DAMS IN CASE OF WATER SPILL OVER A DAM CREST BY USING FAULT TREE ANALYSIS

    Directory of Open Access Journals (Sweden)

    Kuznetsov Dmitriy Viktorovich

    2016-04-01

    Full Text Available The scenario of a hydrodynamic accident of water flow over a crest of a soil dam is considered by the method of fault tree analysis, for which the basic reasons and controlled diagnostic indicators of an accident have been defined. Logical operators “AND”/”OR” were used for creation of a sequence of logically connected events, leading to an undesired event in the scenario of accident. The scenario of the accident was plotted in case of three basic reasons - an excessive settling of a dam crest, an excess flood, an inoperable spillway, taking into account the sequence of the events’ development and with observance of the necessary conditions leading to an accident. “Technical” reasons were observed in the present scenario, force majeure events were not considered. The provided scenario of the accident consists of two branches of events’ development: the left one that depends on an upstream level, and the right one that depends on settling of a dam crest. In each of the considered events an accident “the water spill over a crest of a soil dam” is possible only in case of execution of two different conditions at the same time, i.e. in case of an appropriate upstream level and the appropriate mark of a crest of a soil dam. The conditions of the accident are defined by diagnostic indices - the upstream level and settling of a dam crest, which at the same time are safety criteria of the hydraulic structure for soil dams. They allow defining the technical condition of the construction. Four possible technical conditions are suggested for the definition of technical statuses - normative, operable, limited operable, abnormal. Criteria of safety are the boundaries of the state: for loading and impact - it is the upstream level, for geometrical compliance of the construction - it is a dam crest mark.

  9. Analysis of Japanese radionuclide monitoring data of food before and after the Fukushima nuclear accident.

    Science.gov (United States)

    Merz, Stefan; Shozugawa, Katsumi; Steinhauser, Georg

    2015-03-03

    In an unprecedented food monitoring campaign for radionuclides, the Japanese government took action to secure food safety after the Fukushima nuclear accident (Mar. 11, 2011). In this work we analyze a part of the immense data set, in particular radiocesium contaminations in food from the first year after the accident. Activity concentrations in vegetables peaked immediately after the campaign had commenced, but they decreased quickly, so that by early summer 2011 only a few samples exceeded the regulatory limits. Later, accumulating mushrooms and dried produce led to several exceedances of the limits again. Monitoring of meat started with significant delay, especially outside Fukushima prefecture. After a buildup period, contamination levels of meat peaked by July 2011 (beef). Levels then decreased quickly, but peaked again in September 2011, which was primarily due to boar meat (a known accumulator of radiocesium). Tap water was less contaminated; any restrictions for tap water were canceled by April 1, 2011. Pre-Fukushima (137)Cs and (90)Sr levels (resulting from atmospheric nuclear explosions) in food were typically lower than 0.5 Bq/kg, whereby meat was typically higher in (137)Cs and vegetarian produce was usually higher in (90)Sr. The correlation of background radiostrontium and radiocesium indicated that the regulatory assumption after the Fukushima accident of a maximum activity of (90)Sr being 10% of the respective (137)Cs concentrations may soon be at risk, as the (90)Sr/(137)Cs ratio increases with time. This should be taken into account for the current Japanese food policy as the current regulation will soon underestimate the (90)Sr content of Japanese foods.

  10. SACO-1: a fast-running LMFBR accident-analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  11. Speciation analysis of organotin compounds in lard poisoning accident in Jiangxi Province, China

    Institute of Scientific and Technical Information of China (English)

    江桂斌; 周群芳; 何滨; 刘稷燕

    2000-01-01

    Samples including organotin contaminated lard, urine, blood, and main organs of the poisoned bodies were collected in the incident which took place in Longnan and Dingnan counties, Jiangxi Province, China around the new year’s day of 1999. The organotin compounds in these samples were identified and determined by gas chromatography-flame photometric detector (GC-FPD), gas chromatography-mass spectroscopy (GC-MS) and inductively coupled plasma-mass spectroscopy (ICP-MS). Experiments confirmed that tri- and dimethyltin are the main components that caused the poisoning accident. Monomethyltin, dioctyltin and inorganic tin were also found in several samples.

  12. Analysis on the dispersion of radioactive materials in marine environment after the Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Min, B.; Youm, M.K.; Lee, B.G.; Suh, K.S. [Korea Atomic Energy Research Institute (Korea, Republic of); Raul, P. [Universidad de Sevilla (Spain)

    2014-07-01

    Radioactive materials were released to the atmosphere and ocean due to the accident at the Fukushima Daiichi Nuclear Power Plant (NPP) in March 2011. Marine environment was contaminated by the aeolian fallout and direct release to the sea, especially the radioactive materials were entered in the sea from atmospheric deposition from March 12-30. It was important to evaluate the marine pollution due to the radioactive materials in Pacific Ocean as well as the near fields in the Fukushima Sea. A three-dimensional Lagrangian particle model was used to predict the overall dispersion patterns of the radioactive materials in the global ocean during 2011 to 2021. The spatial domain for the simulations extended from 180 deg. W to 180 deg. E and from 75 deg. S to 75 deg. N. The monthly averaged current data of the 10 years forecast, from JAMSTEC, were used. Numerical simulations were performed to evaluate the distribution of the radionuclides in the ocean with considering directly release and deposition from the atmosphere. Simulated results in the sea water and seabed are compared with the measured data, and atmospheric transport model has been also used to calculate the rates of atmospheric deposition on the sea surface. The clouds of the radioactive materials in surface waters were predicted from 2012 (1 year after accident) to 2021 (10 years after accident) in global ocean. The distributions of the radioactive materials in 2012 showed the rapid movements due to the Kuroshio currents to the eastward direction from Fukushima site. It was predicted that the radioactive clouds reached in the west coast of US after 5 ∼ 6 years from the accident. Comparative results had good agreements in some places over the ocean, but they had a little differences in some locations. The difference between the calculations and measurements are due to the currents and relatively coarse resolutions in the model. The concentrations in dissolved, suspended matters and bottom sediments would be

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

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

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

  14. Speciation analysis of organotin compounds in lard poisoning accident in Jiangxi Province, China

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    Samples including organotin contaminated lard, urine, blood, and main organs of the poisoned bodies were collected in the incident which took place in Longnan and Dingnan counties, Jiangxi Province, China around the new year's day of 1999. The organotin compounds in these samples were identified and determined by gas chromatography-flame photometric detector (GC-FPD), gas chromatography-mass spectroscopy (GC-MS) and inductively coupled plasma- mass spectroscopy (ICP-MS). Experiments confirmed that tri- and dimethyltin are the main components that caused the poisoning accident. Monomethyltin, dioctyltin and inorganic tin were also found in several samples.

  15. Accounting for the cost of occupational accidents

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.

    2004-01-01

    consequences for the company. This, however, presents some challenges due to the current set up of many management accounting systems. The paper explores these issues in the context of the Systematic Accident Cost Analysis (SACA) project, which was carried out during 2001 by The Aarhus School of Business...... and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. It focused on developing and testing a method for the evaluation of the occupational costs and how this might be linked to management accounting and control systems....

  16. Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

    2016-06-15

    Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.

  17. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  18. 40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Internet access to certain off-site... DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off... elements in the risk management plan database available on the Internet: (a) The concentration of...

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

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

    1995-09-01

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

  20. Clinical analysis of 28 children suffering from intracranial hematoma and multiple injuries following traffic accidents

    Institute of Scientific and Technical Information of China (English)

    李江山; 程成; 江勇豪

    2004-01-01

    Objective: To evaluate the result of diagnosis and treatment of intracranial hematoma and multiple injuries caused by road traffic accidents. Methods: Twenty-eight patients, aged from 1 to 14 years, receiving craniotomy and other surgical treatments were retrospectively reviewed. Results: Among the 28 cases, 23 cured with the recovery rate of 82.3%, 2 had a sequel of moderate disability, and 3 died from severe brain injury, hemorrhagic shock, and other visceral complications. The clinical sympotoms and signs were severe and perplexing. The major characters included: severe head injury, usually combined by multiple injuries, and easy of access to missed diagnosis and misdiagnosis. Conclusions: The occurrence of infection is high after traffic accidents as a result of depression of humoral and cellular immunity, long-term bed rest, and fractures of limbs. Hence, on the basis of maintaining vital signs, the management of primary wound is essential to reduce infection and underlying death. In addition to the management of brain injury, concurrent injuries should also be highlighted so as to reach a good result for their patients.

  1. Analysis of Severe Accident for the SFP under the Condition of Drainage using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jung-Min; Pack, Jae-Woo [Jeju National University, Jeju (Korea, Republic of)

    2015-10-15

    This study aims to analyze the effect of a LOCA of the spent fuel pool. We use the MECORE 1.8.6 code to compute the variation of the fuel cladding temperature after a completer loss of the cooling water in the spent fuel pool. A loss of coolant accident in a typical spent fuel pool has been simulated using the MELCOR 1.8.6 code to see the variation of key parameters such as the oxygen concentration in the fuel assembly region and the cladding temperature. In a commercial nuclear power plant, highly radioactive spent fuel assemblies unloaded from the nuclear reactor core are typically stored for a period of time in the spent fuel pool to reduce the radioactivity. The spent fuel assemblies are usually placed in long square racks. It is known that in the progress of the Fukushima nuclear power plant accident, the cooling water in the spent fuel storage was completely lost and the fuel was heated up and damaged. The simulation result shows that the cladding temperature exceeds the rupture temperature in most of the fuel rods and some part of the fuel rods suffers melting of the cladding.

  2. Factors Associated with Fatal Occupational Accidents among Mexican Workers: A National Analysis

    Science.gov (United States)

    Gonzalez-Delgado, Mery; Gómez-Dantés, Héctor; Fernández-Niño, Julián Alfredo; Robles, Eduardo; Borja, Víctor H.; Aguilar, Miriam

    2015-01-01

    Objective To identify the factors associated with fatal occupational injuries in Mexico in 2012 among workers affiliated with the Mexican Social Security Institute. Methods Analysis of secondary data using information from the National Occupational Risk Information System, with the consequence of the occupational injury (fatal versus non-fatal) as the response variable. The analysis included 406,222 non-fatal and 1,140 fatal injuries from 2012. The factors associated with the lethality of the injury were identified using a logistic regression model with the Firth approach. Results Being male (OR=5.86; CI95%: 4.22-8.14), age (OR=1.04; CI95%: 1.03-1.06), employed in the position for 1 to 10 years (versus less than 1 year) (OR=1.37; CI95%: 1.15-1.63), working as a facilities or machine operator or assembler (OR: 3.28; CI95%: 2.12- 5.07) and being a worker without qualifications (OR=1.96; CI95%: 1.18-3.24) (versus an office worker) were associated with fatality in the event of an injury. Additionally, companies classified as maximum risk (OR=1.90; CI 95%: 1.38-2.62), workplace conditions (OR=7.15; CI95%: 3.63-14.10) and factors related to the work environment (OR=9.18; CI95%:4.36-19.33) were identified as risk factors for fatality in the event of an occupational injury. Conclusions Fatality in the event of an occupational injury is associated with factors related to sociodemographics (age, sex and occupation), the work environment and workplace conditions. Worker protection policies should be created for groups with a higher risk of fatal occupational injuries in Mexico. PMID:25790063

  3. Safety accident analysis for second phase projects of Shenzhen metro%深圳地铁二期工程建设期安全事故分析

    Institute of Scientific and Technical Information of China (English)

    杨晨; 张佐汉

    2013-01-01

    By the statistics analysis of safety accidents in Shenzhen Subway Phase Ⅱ Construction,this paper detected that collapse accidents were the accidents most likely happening during construction,and the accidents were prone to happen in the process of foundation pit engineering,shield propelling, underground tunneling and high-formwork construction. By combining with the main methods of subway construction process,this paper showed that auxiliary methods and the temporary bearing structures were the basic reasons for the accident,and put forward the key points of the accident prevention through analysis of the engineering accidents and the construction management experience,which could provide the security management experience for the next subway construction safety.%  通过统计与分析深圳市地铁二期工程在建设期所发生的安全事故,发现坍塌是地铁工程建设期间的多发事故,地铁工程在基坑工程、盾构推进、隧道暗挖和高支模施工的过程中易发生安全事故。结合地铁建设过程中的主要工法,通过进一步分析,发现对辅助工法施工重视不够,临时承重结构施工时抢工期常常是引发事故的重要原因。结合工程事故的分析及建设管理经验,提出事故预防的重点,以期为下阶段大规模的地铁工程建设活动提供安全管理经验。

  4. Analysis of Hydrogen Risk Mitigation System for Severe Accidents of EU-APR1400 Using MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mun Soo; Suh, Jung Soo; Bae, Byoung Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    According to the EUR (European Utility Requirements for LWR Nuclear Power Plants), it is mandatory that the HMS (Hydrogen Mitigation System) of the Eu-APR1400 should be equipped with a passive or automatic hydrogen control system. Considering this requirement, a PAR (Passive Autocatalytic Recombiner) system was adopted for the HMS of the Eu-APR1400. This passive HMS should be evaluated carefully in order to ensure that the HMS has adequate capacity to control hydrogen concentrations during severe accident conditions and to show that the system can satisfy the design requirements of the EUR. In this paper, analyses were carried out to examine the effectiveness of the HMS incorporated into the Eu- APR1400 design. These analyses were performed using the MAAP (Modular Accident Analysis Program) 4 code. in order to identify whether the HMS could control the average hydrogen concentrations in the containment, such that the concentration would not exceed 10 percent by volume: the analyses also considered whether there was the possibility of inadvertent hydrogen combustion in such processes as FA (Flame Acceleration) and DDT (Deflagration to Detonation Transition)

  5. Demonstration sensitivity analysis for RADTRAN III

    Energy Technology Data Exchange (ETDEWEB)

    Neuhauser, K S; Reardon, P C

    1986-10-01

    A demonstration sensitivity analysis was performed to: quantify the relative importance of 37 variables to the total incident free dose; assess the elasticity of seven dose subgroups to those same variables; develop density distributions for accident dose to combinations of accident data under wide-ranging variations; show the relationship between accident consequences and probabilities of occurrence; and develop limits for the variability of probability consequence curves.

  6. The Accident Simulation Analysis of LNG Tank%LNG储罐事故模拟分析

    Institute of Scientific and Technical Information of China (English)

    要栋梁

    2014-01-01

    Based on the software SAFETI coming from Norway Veritas, accident simulation of a liquefied natural gas storage tanks are selected. Leakage and diffusion, tank fire and explosion model was used to simulate the damage scope and impact area. It can pro-vide a theoretical basis for the emergency rescue and disaster preven-tion and mitigation.%本文应用挪威船级社的SFETI软件对某液化天然气储罐进行事故模拟,分别选用泄漏扩散、池火灾和爆炸模型进行模拟计算,得出不同事故的伤害范围和影响区域,为事故救援和防灾减灾提供了理论依据。

  7. Development of auditing technology for accident analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Kim, H. C.; Bae, K. H.; Lee, Y. J.; Chung, Y. J.; Jeong, J. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-06-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement. Well known PIRT methodology has been applied to identify model improvement items. As a part of PIRT process, the identification of SMART-P system and compenent has been performed. The scenario of each key accident and phenonema have been identified. To identify SMART-P thermal-hydraulic characteristics, preliminary calculation has been performed and identify the applicability and inprovement items of current auditing code, RELAP5.

  8. Pre-Study of Off-site Consequence Analysis in Level 3 PSA of Wolsong Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Jik; Yang, Ho-Chang; Choi, Seong-Soo [ACT, Daejeon (Korea, Republic of)

    2015-10-15

    In order to perform level 3 PSA, MACCS II (MELCOR Accident Consequence Code System 2) is needed. MACCS II is used in PSA for plants in order to evaluate population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event in CANDU-6 plants is evaluated population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event has been evaluated by using Level 1 PSA result and Level 2 PSA result(ISSAC) and MACCS II. As a result, We are obtained the following conclusion. - Early maximum early fatalities is 5.35E+02 equal to latent maximum early fatalities.(99.5%) - Early and latent maximum cancer fatalities are 2.33E+03 and 1.11E+04, respectively. (99.5%) - Early and latent maximum population doses are 1.25 and 5.00 person-rem/yr, respectively. (99.5%) Other study has shown that MACCS II was performed evaluation for Wolsong NPP. Small Break Loss of Coolant Accident(SBLOCA) event is selected by other study. The results of early and cancer fatalities applied similar assumption were 3.02E+00 and 1.89E+03, respectively. This study's results are higher than other study's result. Because, basis input data is different each studies, and event frequency are different (This study : 2.10E-07/ Other study : 4.93E-09)

  9. Shipping container response to severe highway and railway accident conditions: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

  10. 基于HFACS的海上交通事故原因系统分析%Systemic analysis on cause of marine traffic accidents based on HFACS

    Institute of Scientific and Technical Information of China (English)

    张欣欣; 轩少永; 席永涛; 胡甚平

    2012-01-01

    为定量研究海上交通事故人失误致因因素,分析各种人失误因素对事故的影响程度,从而达到控制人失误事故的最终目的.在引入人的不安全行为分类框架和“人-机-环境”系统的基础上,运用人因素分析与分类系统(Human Factors Analysis and Classification System,HFACS),提出海上交通事故人失误分析与分类系统(Human Error Analysis and Classification System for Marine Traffic Accident,HEACS-MTA),对海上交通事故人失误因素进行分类.运用灰色关联分析法(Grey Relational Analysis,GRA)对事故形成原因进行定量分析,得出管理因素是事故的根本原因.导致事故发生的人失误因素依次为不安全行为的前提条件、不安全的监督、不安全行为和组织影响.%To research the causation factors of human errors in marine traffic accidents quantitatively, the influencing degree of the causation factors which lead to human errors is analyzed, so as to control accidents caused by human errors. On the base of introducing the classification framework of human' s unsafe behavior and the " man-machine-environment" system, the Human Factors Analysis and Classification System ( HFACS) is used and the Human Error Analysis and Classification System for Marine Traffic Accident (HEACS-MTA) is proposed to classify human errors in marine traffic accidents. The Grey Relational Analysis ( GRA) is used to analyze the accident causes quantitatively, and the conclusion is made that the management factor is the root cause of the accidents. The order of the main human error factors which lead to accident is precondition for unsafe acts, unsafe supervision, unsafe acts and organizational influences.

  11. Tools for improving safety management in the Norwegian Fishing Fleet occupational accidents analysis period of 1998-2006.

    Science.gov (United States)

    Aasjord, Halvard L

    2006-01-01

    Reporting of human accidents in the Norwegian Fishing Fleet has always been very difficult because there has been no tradition in making reports on all types of working accidents among fishermen, if the accident does not seem to be very serious or there is no economical incentive to report. Therefore reports are only written when the accidents are serious or if the fisherman is reported sick. Reports about an accident are sent to the insurance company, but another report should also be sent to the Norwegian Maritime Directorate (NMD). Comparing of data from one former insurance company and NMD shows that the real numbers of injuries or serious accidents among Norwegian fishermen could be up to two times more than the numbers reported to NMD. Special analyses of 1690 accidents from the so called PUS-database (NMD) for the period 1998-2002, show that the calculated risk was 23.6 accidents per 1000 man-years. This is quite a high risk level, and most of the accidents in the fishing fleet were rather serious. The calculated risks are highest for fishermen on board the deep sea fleet of trawlers (28.6 accidents per 1000 man-years) and also on the deep sea fleet of purse seiners (28.9 accidents per 1000 man-years). Fatal accidents over a longer period of 51.5 years from 1955 to 2006 are also roughly analysed. These data from SINTEF's own database show that the numbers of fatal accidents have been decreasing over this long period, except for the two periods 1980-84 and 1990-94 where we had some casualties with total losses of larger vessels with the loss of most of the crew, but also many others typical work accidents on smaller vessels. The total numbers of registered Norwegian fishermen and also the numbers of man-years have been drastically reduced over the 51.5 years from 1955 to 2006. The risks of fatal accidents have been very steady over time at a high level, although there has been a marked risk reduction since 1990-94. For the last 8.5-year period of January 1998

  12. Analysis and prevention on explosion accident of sulfuric acid tanks%硫酸储罐爆炸事故分析及预防

    Institute of Scientific and Technical Information of China (English)

    张启波; 袁凤丽; 王清

    2012-01-01

    介绍了山东省某化工有限公司苯胺厂的工人在废硫酸罐顶部焊接管线时发生的一起废硫酸罐爆炸事故.通过对事故发生经过及现场情况的调查分析,找出了导致事故发生的原因,由于废硫酸罐耐酸瓷瓦破损,废硫酸渗漏与罐体接触反应产生的氢气,与由苯-稀硫酸萃取分离器串入废硫酸罐的苯或硝基苯蒸气及罐内空气混合形成爆炸性混合物,遇到因违章操作产生的明火、高温发生爆炸.通过对这次事故的详细描述、分析,在吸取事故教训的基础上,提出了相应的预防措施,为预防类似事故的发生提供参考.%In February of 2007, an accident happened in a chemical Co. , LTD aniline factory in Shandong Province. When welding the pipe of the waste sulphuric acid can on the top of the waste sulphuric acid , the tank exploded suddenly. Based on a large number of investigation and analysis on the accident process and the scene of the accident, the reasons that caused the accident were found. The damage of waste sulphuric acid porcelain tile leaded to waste sulphuric acid leakage and generated hydrogen by reaction with the tank body. Benzene and nitrobenzene steam gone by benzene-dilute sulphuric acid extraction separator strung into the workers were on waste sulphuric acid tank. Hydrogen, benzene and nitrobenzene steam and air in the tank were mixed into explosive mixtures. These explosive mixtures encountered open flame and high temperature resulted from workers'illegal operations and then exploded, which caused the tragic accident. Facing the accident, and taking it as a warning for future, through the detailed description and analysis of the accident and in the draw lessons from the accident, the corresponding prevention measures were put forward, in order to prevent the occurrence of similar accidents.

  13. Fault tree analysis of fire and explosion accidents for dual fuel (diesel/natural gas) ship engine rooms

    Science.gov (United States)

    Guan, Yifeng; Zhao, Jie; Shi, Tengfei; Zhu, Peipei

    2016-09-01

    In recent years, China's increased interest in environmental protection has led to a promotion of energy-efficient dual fuel (diesel/natural gas) ships in Chinese inland rivers. A natural gas as ship fuel may pose dangers of fire and explosion if a gas leak occurs. If explosions or fires occur in the engine rooms of a ship, heavy damage and losses will be incurred. In this paper, a fault tree model is presented that considers both fires and explosions in a dual fuel ship; in this model, dual fuel engine rooms are the top events. All the basic events along with the minimum cut sets are obtained through the analysis. The primary factors that affect accidents involving fires and explosions are determined by calculating the degree of structure importance of the basic events. According to these results, corresponding measures are proposed to ensure and improve the safety and reliability of Chinese inland dual fuel ships.

  14. Fault Tree Analysis of Fire and Explosion Accidents for Dual Fuel (Diesel/Natural Gas) Ship Engine Rooms

    Institute of Scientific and Technical Information of China (English)

    Yifeng Guan; Jie Zhao; Tengfei Shiand Peipei Zhu

    2016-01-01

    In recent years, China’s increased interest in environmental protection has led to a promotion of energy-efficient dual fuel (diesel/natural gas) ships in Chinese inland rivers. A natural gas as ship fuel may pose dangers of fire and explosion if a gas leak occurs. If explosions or fires occur in the engine rooms of a ship, heavy damage and losses will be incurred. In this paper, a fault tree model is presented that considers both fires and explosions in a dual fuel ship;in this model, dual fuel engine rooms are the top events. All the basic events along with the minimum cut sets are obtained through the analysis.The primary factors that affect accidents involving fires and explosions are determined by calculating the degree of structure importance of the basic events.According to these results, corresponding measures are proposed to ensure and improve the safety and reliability of Chinese inland dual fuel ships.

  15. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  16. Optimized electricity expansions with external costs internalized and risk of severe accidents as a new criterion in the decision analysis

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Estrada S, G. J., E-mail: cmcm@fi-b.unam.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2011-11-15

    The external cost of severe accidents was incorporated as a new element for the assessment of energy technologies in the expansion plans of the Mexican electric generating system. Optimizations of the electric expansions were made by internalizing the external cost into the objective function of the WASP-IV model as a variable cost, and these expansions were compared with the expansion plans that did not internalize them. Average external costs reported by the Extern E Project were used for each type of technology and were added to the variable component of operation and maintenance cost in the study cases in which the externalises were internalized. Special attention was paid to study the convenience of including nuclear energy in the generating mix. The comparative assessment of six expansion plans was made by means of the Position Vector of Minimum Regret Analysis (PVMRA) decision analysis tool. The expansion plans were ranked according to seven decision criteria which consider internal costs, economical impact associated with incremental fuel prices, diversity, external costs, foreign capital fraction, carbon-free fraction, and external costs of severe accidents. A set of data for the calculation of the last criterion was obtained from a Report of the European Commission. We found that with the external costs included in the optimization process of WASP-IV, better electric expansion plans, with lower total (internal + external) generating costs, were found. On the other hand, the plans which included the participation of nuclear power plants were in general relatively more attractive than the plans that did not. (Author)

  17. Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

    2015-05-15

    The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

  18. Self-reported accidents

    DEFF Research Database (Denmark)

    Møller, Katrine Meltofte; Andersen, Camilla Sloth

    2016-01-01

    The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

  19. "Murder-suicide" or "murder-accident"? Difficulties with the analysis of cases.

    Science.gov (United States)

    Byard, Roger W; Veldhoen, David; Kobus, Hilton; Heath, Karen

    2010-09-01

    Homicide where a perpetrator is found dead adjacent to the victim usually represents murder-suicide. Two incidents are reported to demonstrate characteristic features in one, and alternative features in the other, that indicate differences in the manner of death. (i) A 37-year-old mother was found dead in a burnt out house with her two young sons in an adjacent bedroom. Deaths were due to incineration and inhalation of products of combustion. (ii) A 39-year-old woman was found stabbed to death in a burnt out house with her 39-year-old de facto partner deceased from the combined effects of incineration and inhalation of products of combustion. The first incident represented a typical murder-suicide, however, in the second incident, the perpetrator had tried to escape through a window and had then sought refuge in a bathroom under a running shower. Murder-accident rather than murder-suicide may therefore be a more accurate designation for such cases.

  20. An analysis of Japan radiation protection measurements after the Fukushima nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Arlene A. dos; Souza-Santos, Denison; Melo, Dunstana R. de; Hunt, John G.; Juliao, Ligia M.Q C.; Conti, Luiz F.C.; Pires do Rio, Monica A.; Reis, Rocio G., E-mail: arlene@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    On March 11th 2011, Japan was struck by a devastating earthquake followed by a tsunami wave that took the lives of thousands and started a major nuclear accident in the Fukushima Dai-ichi power plant complex. Right from the beginning, the information published by the Japanese government and by the International Atomic Energy Agency (IAEA) was followed by a team of experts at the Institute for Radiation Protection and Dosimetry (IRD) of Brazil. Radiation monitoring data, such as radionuclide activity concentration in water and food, ambient dose rate and fallout concentration in specific cities have been compiled and analyzed, with emphasis on dose limits established by Brazilian regulatory authority. A computer code for dose assessment, developed at the IRD and based upon the IAEA documents TECDOC-1162 and TECDOC-955, was used to assess the doses due to intakes of radionuclides and external exposure for individuals of different age groups. The IAEA model predictions for the ambient dose rates, when the fallout is known, are compared with the measured values in different cities. The Japanese recommendations for evacuation, sheltering and restriction of food and water consumption are evaluated with regards to the Brazilian limits defined in the CNEN NN 3.01 standard. (author)

  1. 尿素合成塔内化学爆炸机理分析%Theoretical analysis on the urea reactor explosion accidents

    Institute of Scientific and Technical Information of China (English)

    李冬; 陶刚; 张礼敬; 许丹枫

    2011-01-01

    分析了尿素合成塔发生化学爆炸的可能性.利用经验公式计算出塔内上部空间气相在正常生产和非正常生产状态下的爆炸极限范围,并计算出非正常生产状态下气相化学爆炸产生的压力和能量范围.计算结果与现场数据吻合.爆炸所产生的压力作用于液面上,根据静压传递原理,尿液内各点与外压相等,超过尿塔正常工作压力,并导致破裂.%This paper intends to take as its two case study samples the two serious urea reactor explosion accidents that took place at home, which brought about heavy casualty and economic damage of similar nature. Thus, our analysis has been done in terms of three key elements of chemical explosions and their aftermath consequences. The analyses of the explosion process indicate that the flammable gases were found including hydrogen, ammonia and methane mixed with oxygen and inert gases, such as nitrogen, vapor and carbon dioxide in the gas phase of urea reactor. The impact friction, adiabatic compression, static and spontaneous combustion of such gases were found to be the ignition sources as the result of the analysis. Empirical equations were also used to confirm the explosion limit under the condition of high temperatures and high pressures.Moreover, conditions of inert gas presence was also taken into account. Everything considered, the explosion limit of mixed gases was found to be 11.27% - 78.94% under the normal production conditions. And the practical mixed gas concentration was found to be 92.19%, far above the explosion maximum limit. On the other hand, if the temperature decreased and the pressure remained in the maintenance range, and if the practical mixed gas concentration was within the explosion limit, it would also be possible to cause chemical explosions. When the urea reactor were shut down with its pressure maintained and the thermos level dropping, the volume of the gas phase would be likely to approach 3.7 m3. The

  2. Exploring Environmental Effects of Accidents During Marine Transport of Dangerous Goods by Use of Accident Descriptions

    DEFF Research Database (Denmark)

    Rømer, Hans Gottberg; Haastrup, P.; Petersen, H J Styhr

    1996-01-01

    On the basis of 1776 descriptions of water transport accidents involving dangerous goods, environmental problems in connection with releases of this kind are described and discussed. It was found that most detailed descriptions of environmental consequences concerned oil accidents, although most...

  3. Spatiotemporal Analysis for Wildlife-Vehicle Based on Accident Statistics of the County Straubing-Bogen in Lower Bavaria

    Science.gov (United States)

    Pagany, R.; Dorner, W.

    2016-06-01

    During the last years the numbers of wildlife-vehicle-collisions (WVC) in Bavaria increased considerably. Despite the statistical registration of WVC and preventive measures at areas of risk along the roads, the number of such accidents could not be contained. Using geospatial analysis on WVC data of the last five years for county Straubing-Bogen, Bavaria, a small-scale methodology was found to analyse the risk of WVC along the roads in the investigated area. Various indicators were examined, which may be related to WVC. The risk depends on the time of the day and year which shows correlations in turn to the traffic density and wildlife population. Additionally the location of the collision depends on the species and on different environmental parameters. Accidents seem to correlate with the land use left and right of the street. Land use data and current vegetation were derived from remote sensing data, providing information of the general land use, also considering the vegetation period. For this a number of hot spots was selected to identify potential dependencies between land use, vegetation and season. First results from these hotspots show, that WVCs do not only depend on land use, but may show a correlation with the vegetation period. With regard to agriculture and seasonal as well as annual changes this indicates that warnings will fail due to their static character in contrast to the dynamic situation of land use and resulting risk for WVCs. This shows that there is a demand for remote sensing data with a high spatial and temporal resolution as well as a methodology to derive WVC warnings considering land use and vegetation. With remote sensing data, it could become possible to classify land use and calculate risk levels for WVC. Additional parameters, derived from remote sensed data that could be considered are relief and crops as well as other parameters such as ponds, natural and infrastructural barriers that could be related to animal behaviour and

  4. Analysis of Individual and Environmental Factors for Road Traffic Accidents in Sirjan-Bandarabbas Road between 2010 and 2011, Iran

    Directory of Open Access Journals (Sweden)

    Ghorbanali Mohammadi

    2016-10-01

    Full Text Available Sirjan -Bandarabbas road is one of the important commercial roads in Iran and for Sirjan’s area situation and relevance between Sirjan and other states in Iran so high percentage of goods that forwarded from Bandarabbas to other states transit from Sirjan .Therefore this road is as one important transition road and traffic road too .This study analyzed road traffic accidents were occurred between 2010 and 201in Sirjan- Bandarabbas road. Individual and demographic factors include Time of accidents, Drivers age, time of the days, seat belt and safety laws, Guilty vehicle, Mode of accident and education Level. Time of day analyses suggested that the highest percentage of road traffic injuries occurred in the time group between 12-18 hours. Drivers with the age group of 36-50 had more involvement in death accidents. The findings of this study also revealed that most of the collisions was front to back and front to side. Female drivers were found to be generally safer drivers than their male counterparts; male drivers had a higher involvement rate in road traffic accidents. This study indicated that Observe safety laws, Guilty vehicle and Mode of accidents have a meaningful relationship with Type of accidents in road traffic accidents in Sirjan Bandarabbas road.

  5. Modeling and sensitivity analysis of transport and deposition of radionuclides from the Fukushima Daiichi accident

    Directory of Open Access Journals (Sweden)

    X. Hu

    2014-01-01

    Full Text Available The atmospheric transport and ground deposition of radioactive isotopes 131I and 137Cs during and after the Fukushima Daiichi Nuclear Power Plant (FDNPP accident (March 2011 are investigated using the Weather Research and Forecasting/Chemistry (WRF/Chem model. The aim is to assess the skill of WRF in simulating these processes and the sensitivity of the model's performance to various parameterizations of unresolved physics. The WRF/Chem model is first upgraded by implementing a radioactive decay term into the advection-diffusion solver and adding three parameterizations for dry deposition and two parameterizations for wet deposition. Different microphysics and horizontal turbulent diffusion schemes are then tested for their ability to reproduce observed meteorological conditions. Subsequently, the influence on the simulated transport and deposition of the characteristics of the emission source, including the emission rate, the gas partitioning of 131I and the size distribution of 137Cs, is examined. The results show that the model can predict the wind fields and rainfall realistically. The ground deposition of the radionuclides can also potentially be captured well but it is very sensitive to the emission characterization. It is found that the total deposition is most influenced by the emission rate for both 131I and 137Cs; while it is less sensitive to the dry deposition parameterizations. Moreover, for 131I, the deposition is also sensitive to the microphysics schemes, the horizontal diffusion schemes, gas partitioning and wet deposition parameterizations; while for 137Cs, the deposition is very sensitive to the microphysics schemes and wet deposition parameterizations, and it is also sensitive to the horizontal diffusion schemes and the size distribution.

  6. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  7. Road characteristics and bicycle accidents.

    Science.gov (United States)

    Nyberg, P; Björnstig, U; Bygren, L O

    1996-12-01

    In Umeå, Sweden, defects in the physical road surface contributed to nearly half of the single bicycle accidents. The total social cost of these injuries to people amount to at least SEK 20 million (SEK 60,000 or about USD 8,500 per accident), which corresponds to the estimated loss of "eight life equivalents a year". Improved winter maintenance seems to have the greatest injury prevention potential and would probably reduce the number of injuries considerably, whereas improved road quality and modification of kerbs would reduce the most severe injuries. A local traffic safety program should try to prevent road accidents instead of handling the consequences of them. In accordance with Parliament decisions on traffic we would like to see increased investment in measures favoring bicycle traffic, where cycling is seen as a solution, not as a problem.

  8. Systematic approach for assessment of accident risks in chemical and nuclear processing; Abordagem sistematica para avaliacao de riscos de acidentes em instalacoes de processamento quimico e nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Senne Junior, Murillo

    2003-07-15

    The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of

  9. Calculation of Departure from Nucleate Boiling Ratio (DNBR) minimum for accident analysis of main steam line break at Angra-1; Calculo do minimo DNBR para analise do acidente de ruptura da linha principal de vapor em Angra-1

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Marcio Dornellas [ELETROBRAS Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil). E-mail: mdorne@eletronuclear.gov.br

    2000-07-01

    The maintenance costs, the operational problems and the failures possibilities of the boron injection system, composed by pumps, valves, heated lines and the boron injection tank, make this tank removal or the boron concentration reduction advisable for Angra 1 Power Plant. The main accident from chapter XV of the final safety analysis report affected by this modification is the main steam line break. It is necessary the interaction of the areas of Accidents and Transients Analysis (RETRAN 02/Mod 5.1 code), Neutronics (APA System) and Thermohydraulics (COBRA IIIC/MIT) to analyse this accident. The present Angra 1 boron concentration is 20000 ppm and it could be reduced to 2000 ppm as a result of the present study. The Departure from Nucleate Boiling Ratio (DNBR) is the restrictive parameter of this accident, which is calculated from the initials and boundary conditions obtained from the Transients and Accidents Analysis and Neutronics areas. (author)

  10. Analysis on Causes of Electrical Shock Accident of Fountain%喷水池电击事故起因简析

    Institute of Scientific and Technical Information of China (English)

    王厚余

    2012-01-01

    Electrical shock accident of fountain occurs frequently in China. Electrical equipments with a voltage of 220 V and above should be used for fountain and over-high voltage will result in electrical shock accident underwater. The author believes that different from that above ground, the electrical shock accident of fountain is mainly caused by underground voltage gradient or electric field. Therefore, relevant analysis is conducted and key points to prevent electrical shock accident during electrical design under the current conditions in China are presented.%我国喷水池电击事故屡有发生。喷水池内电气设备需采用220V及以上电压,过高电压在水下很易引发电击事故。作者认为喷水池内电击事故的发生与地面上不同,它主要由水下电压梯度或电场引起。故对此进行分析,并提出在我国现时条件下喷水池电气设计中防电击的要点。

  11. Analysis of design strategies for mitigating the consequences of lithium fire within containment of controlled thermonuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dube, D A; Kazimi, M S

    1978-07-01

    A lithium combustion model (LITFIRE) was developed to describe the physical and chemical processes which occur during a hypothetical lithium spill and fire. The model was used to study the effectiveness of various design strategies for mitigating the consequences of lithium fire, using the UWMAK-III features as a reference design. Calculations show that without any special fire protection measures, the containment may reach pressures of up to 32 psig when one coolant loop is spilled inside the reactor building. Temperatures as high as 2000/sup 0/F would also be experienced by some of the containment structures. These consequences were found to diminish greatly by the incorporation of a number of design strategies including initially subatmospheric containment pressures, enhanced structural surface heat removal capability, initially low oxygen concentrations, and active post-accident cooling of the containment gas. The EBTR modular design was found to limit the consequences of a lithium spill, and hence offers a potential safety advantage. Calculations of the maximum flame temperature resulting from lithium fire indicate that none of the radioactive first wall materials under consideration would vaporize, and only a few could possibly melt.

  12. CFD Simulation of a fall accident of a fuel element in pool This project aims at calculating the speed ratio of impact-fall height for a PWR fuel element falling freely in the fuel pool; Simulacion CFD de un accidente de caida de un elemento combustible en piscina

    Energy Technology Data Exchange (ETDEWEB)

    Montoro Garcia, B.; Corpa Masa, R.; Jimenez-Reja, C.

    2014-07-01

    It is intended to provide a methodology of analysis more realistic this accident.que referred to in calculations of the license that requires fuel catastrophic break regardless of the height of the fall, with the consequent release of inventory analysers. Accidents that occurred in the past indicate that this hypothesis could be too conservative. (Author)

  13. Forced displacement in Yugoslavia: a meta-analysis of psychological consequences and their moderators.

    Science.gov (United States)

    Porter, M; Haslam, N

    2001-10-01

    A meta-analysis was conducted to synthesize what is known about differences in mental health between refugees and nonrefugees from the former Yugoslavia. The analysis focused on moderating effects of a variety of enduring, contextual stressors. Results indicated that refugees suffer significantly more mental health impairment than nonrefugees. The psychological consequences of forced displacement were found to vary significantly as a function of chronic stressors (e.g., locus of displacement and type of accommodation in exile) and were also associated with otherfactors (e.g., degree of war exposure in the nondisplaced groups, participant age, and time of data collection as reflected in year of publication). Implications for the study of refugee mental health are discussed.

  14. Consequences of radioactive releases into the sea resulting from the accident at the Fukushima Dai-ichi nuclear power plant - Evolution of expert investigation according to the data available

    OpenAIRE

    2012-01-01

    The accident at the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) in March 2011 led to an unprecedented direct input of artificial radioactivity into the marine environment. The Institute for Radioprotection and Nuclear Safety was requested by the French authorities to investigate the radioecological impact of this input, in particular the potential contamination of products of marine origin used for human consumption. This article describes the close link between the responses provided and ...

  15. Quantitative risk analysis of gas explosions in tunnels; probability, effects, and consequences

    NARCIS (Netherlands)

    Weerheijm, J.; Voort, M.M. van der; Verreault, J.; Berg, A.C. van den

    2015-01-01

    Tunnel accidents with transports of combustible liquefied gases may lead to explosions. Depending on the substance involved this can be a Boiling Liquid Expanding Vapour Explosion (BLEVE), a Gas Expansion Explosion (GEE) or a gas explosion. Quantification of the risk of these scenarios is important

  16. Consequence Management Joint Center for Operational Analysis Journal, Volume 11, Issue 1, Winter 2008-2009

    Science.gov (United States)

    2009-01-01

    biological parameters, such as gene activation or chromosomal abnormalities, or on the physical changes of tissues, and can be detected by...current biodosimetry methods for radiation incidents and accidents can be divided into three groups: (1) Cytogenetics a. Dicentric assay b...Fluorescence in situ hybridization (FISH) assay c. Cytokinesis block micronucleus (CBMN) assay d. Premature chromosome condensation (PCC) assay JCOA

  17. Paragliding accidents in remote areas.

    Science.gov (United States)

    Fasching, G; Schippinger, G; Pretscher, R

    1997-08-01

    Paragliding is an increasingly popular hobby, as people try to find new and more adventurous activities. However, there is an increased and inherent danger with this sport. For this reason, as well as the inexperience of many operators, injuries occur frequently. This retrospective study centers on the helicopter rescue of 70 individuals in paragliding accidents. All histories were examined, and 43 patients answered a questionnaire. Nineteen (42%) pilots were injured when taking off, 20 (44%) during the flight, and six (13%) when landing. Routine and experience did not affect the prevalence of accident. Analysis of the causes of accident revealed pilot errors in all but three cases. In 34 rescue operations a landing of the helicopter near the site of the accident was possible. Half of the patients had to be rescued by a cable winch or a long rope fixed to the helicopter. Seven (10%) of the pilots suffered multiple trauma, 38 (54%) had injuries of the lower extremities, and 32 (84%) of them sustained fractures. Injuries to the spine were diagnosed in 34 cases with a fracture rate of 85%. One patient had an incomplete paraplegia. Injuries to the head occurred in 17 patients. No paraglider pilot died. The average hospitalization was 22 days, and average time of working inability was 14 weeks. Fourteen (34%) patients suffered from a permanent damage to their nerves or joints. Forty-three percent of the paragliders continued their sport despite the accident; two of them had another accident. An improved training program is necessary to lower the incidence of paragliding accidents. Optimal equipment to reduce injuries in case of accidents is mandatory. The helicopter emergency physician must perform a careful examination, provide stabilization of airways and circulation, give analgesics, splint fractured extremities, and transport the victim on a vacuum mattress to the appropriate hospital.

  18. An evaluation of spindle-shaft seizure accident sequences for the Schenck Dynamic Balancer

    Energy Technology Data Exchange (ETDEWEB)

    Bott, T.F.; Fischer, S.R.

    1998-11-01

    This study was conducted at the request of the USDOE/AL Dynamic Balancer Project Team to develop a set of representative accident sequences initiated by rapid seizure of the spindle shaft of the Schenck dynamic balancing machine used in the mass properties testing activities in Bay 12-60 at the Pantex Plant. This Balancer is used for balancing reentry vehicles. In addition, the study identified potential causes of possible spindle-shaft seizure leading to a rapid deceleration of the rotating assembly. These accident sequences extend to the point that the reentry vehicle either remains in stable condition on the balancing machine or leaves the machine with some translational and rotational motion. Fault-tree analysis was used to identify possible causes of spindle-shaft seizure, and failure modes and effects analysis identified the results of shearing of different machine components. Cause-consequence diagrams were used to help develop accident sequences resulting from the possible effects of spindle-shaft seizure. To make these accident sequences physically reasonable, the analysts used idealized models of the dynamics of rotating masses. Idealized physical modeling also was used to provide approximate values of accident parameters that lead to branching down different accident progression paths. The exacerbating conditions of balancing machine over-speed and improper assembly of the fixture to the face plate are also addressed.

  19. The Human Factors of an Early Space Accident: Flight 3-65 of the X-15

    Science.gov (United States)

    Barshi, Immanuel; Statler, Irving C.; Orr, Jeb S.

    2016-01-01

    The X-15 was a critical research vehicle in the early days of space flight. On November 15, 1967, the X-15-3 suffered an in-flight breakup. This 191st flight of the X-15 and the 65th flight of this third configuration was the only fatal accident of the X-15 program. This paper presents an analysis, from a human factors perspective, of the events that led up to the accident. The analysis is based on the information contained in the report of the Air Force-NASA Accident Investigation Board (AIB) dated January, 1968. The AIBs analysis addressed, primarily, the events that occurred subsequent to the pilot's taking direct control of the reaction control system. The analysis described here suggests that, rather than events following the pilot's switch to direct control, it was the events preceding the switch that led to the accident. Consequently, the analyses and conclusions regarding the causal factors of, and the contributing factors to, the loss of Flight 3-65 presented here differ from those of the AIB based on the same evidence. Although the accident occurred in 1967, the results of the presented analysis are still relevant today. We present our analysis and discuss its implications for the safety of space operations.

  20. SPATIOTEMPORAL ANALYSIS FOR WILDLIFE-VEHICLE-COLLISIONS BASED ON ACCIDENT STATISTICS OF THE COUNTY STRAUBING-BOGEN IN LOWER BAVARIA

    Directory of Open Access Journals (Sweden)

    R. Pagany

    2016-06-01

    Full Text Available During the last years the numbers of wildlife-vehicle-collisions (WVC in Bavaria increased considerably. Despite the statistical registration of WVC and preventive measures at areas of risk along the roads, the number of such accidents could not be contained. Using geospatial analysis on WVC data of the last five years for county Straubing-Bogen, Bavaria, a small-scale methodology was found to analyse the risk of WVC along the roads in the investigated area. Various indicators were examined, which may be related to WVC. The risk depends on the time of the day and year which shows correlations in turn to the traffic density and wildlife population. Additionally the location of the collision depends on the species and on different environmental parameters. Accidents seem to correlate with the land use left and right of the street. Land use data and current vegetation were derived from remote sensing data, providing information of the general land use, also considering the vegetation period. For this a number of hot spots was selected to identify potential dependencies between land use, vegetation and season. First results from these hotspots show, that WVCs do not only depend on land use, but may show a correlation with the vegetation period. With regard to agriculture and seasonal as well as annual changes this indicates that warnings will fail due to their static character in contrast to the dynamic situation of land use and resulting risk for WVCs. This shows that there is a demand for remote sensing data with a high spatial and temporal resolution as well as a methodology to derive WVC warnings considering land use and vegetation. With remote sensing data, it could become possible to classify land use and calculate risk levels for WVC. Additional parameters, derived from remote sensed data that could be considered are relief and crops as well as other parameters such as ponds, natural and infrastructural barriers that could be related to

  1. Comparative Assessment of Severe Accidents in the Chinese Energy Sector

    Energy Technology Data Exchange (ETDEWEB)

    Hirschberg, S.; Burgherr, P.; Spiekerman, G.; Cazzoli, E.; Vitazek, J.; Cheng, L

    2003-03-01

    This report deals with the comparative assessment of accidents risks characteristic for the various electricity supply options. A reasonably complete picture of the wide spectrum of health, environmental and economic effects associated with various energy systems can only be obtained by considering damages due to normal operation as well as due to accidents. The focus of the present work is on severe accidents, as these are considered controversial. By severe accidents we understand potential or actual accidents that represent a significant risk to people, property and the environment and may lead to large consequences. (author)

  2. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  3. Syntactic accidents in program analysis: on the impact of the CPS transformation

    DEFF Research Database (Denmark)

    Damian, Daniel; Danvy, Olivier

    2003-01-01

    on a direct-style program and on its CPS counterpart. Our proof technique amounts to constructing the CPS counterpart of flow information and of binding times. Our results formalize and confirm a folklore theorem about traditional binding-time analysis, namely that CPS has a positive effect on binding times......We show that a non-duplicating transformation into Continuation-Passing Style (CPS) has no effect on control-flow analysis, a positive effect on binding-time analysis for traditional partial evaluation, and no effect on binding-time analysis for continuation-based partial evaluation: a monovariant...... in source programs, i.e., to the way these programs are written. More reliable program analyses require a better understanding of the effect of syntactic change....

  4. Syntactic Accidents in Program Analysis: On the Impact of the CPS Transformation

    DEFF Research Database (Denmark)

    Daniel, Damian; Danvy, Olivier

    2000-01-01

    on a direct-style program and on its CPS counterpart. Our proof technique amounts to constructing the CPS counterpart of flow information and of binding times. Our results formalize and confirm a folklore theorem about traditional binding-time analysis, namely that CPS has a positive effect on binding times......We show that a non-duplicating transformation into Continuation-Passing Style (CPS) has no effect on control-flow analysis, a positive effect on binding-time analysis for traditional partial evaluation, and no effect on binding-time analysis for continuation-based partial evaluation: a monovariant...... in source programs, i.e., to the way these programs are written. More reliable program analyses require a better understanding of the effect of syntactic change....

  5. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, N.; Khedr, A. [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt); D' Auria, F.D. [Pisa Univ. (Italy). Facolta di Ingegneria

    2015-12-15

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  6. [Accidents in travellers - the hidden epidemic].

    Science.gov (United States)

    Walz, Alexander; Hatz, Christoph

    2013-06-01

    The risk of malaria and other communicable diseases is well addressed in pre-travel advice. Accidents are usually less discussed. Thus, we aimed at assessing accident figures for the Swiss population, based on data of the register from 2004 to 2008 of the largest Swiss accident insurance organization (SUVA). More than 139'000 accidents over 5 years showed that 65 % of the accidents overseas are injuries, and 24 % are caused by poisoning or harm by cold, heat or air pressure. Most accidents happened during leisure activities or sports. More than one third of the non-lethal and more than 50 % of the fatal accidents happened in Asia. More than three-quarters of non-lethal accidents take place in people between 25 and 54 years. One out of 74 insured persons has an accident abroad per year. Despite of many analysis short-comings of the data set with regard to overseas travel, the figures document the underestimated burden of disease caused by accidents abroad and should affect the given pre-health advice.

  7. Validation of GAMMA+ model for Evaluating Heat Transfer of VHTR core in Accident Conditions by CFD analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dongho; Yoon, Sujong; Park, Gooncherl; Cho, Hyoungkyu [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    KAERI has established a plan to demonstrate massive production of hydrogen using a VHTR by the early 2020s. In addition the GAMMA+ code is developed to analyze VHTR thermo-fluid transients at KAERI. One of the candidate reactor designs for VHTR is prismatic modular reactor (PMR), of which reference reactor is the 600MWth GT-MHR. This type of reactor has a passive safety system. During the High Pressure Conduction Cooling (HPCC) or Low Pressure Conduction Cooling (LPCC) accident, the core heats up by decay heat and then starts to cool down by conduction and radiation cooling to the Reactor Cavity Cooling System (RCCS) through the prismatic core. In this mechanism, the solid conduction occurs in graphite and fuel blocks, and the gas conduction and radiation occurs in coolant holes and bypass gaps. It is important to predict conduction and radiation heat transfer in the core for safety analysis. Effective thermal conductivity is derived by Maxwell's far-field methodology Radiation effect is expressed as corresponding conductivity and added to gas conductivity. In this study, ETC model used in GAMMA+ code is validated with the commercial CFD code, CFX-13. In this study, the effective thermal conductivity model of the GAMMA+ was evaluated by comparison of CFD analysis. The CFD analysis was conducted for various numbers and volume fractions of coolant holes and temperatures. Although slight disagreement was shown for the cases run with small number of holes, the result of GAMMA+ model is accurate for the large numbers of holes sufficiently. Since there are 102 coolant holes and 210 fuel holes in a fuel block, it is concluded that GAMMA+ model is proper formula for predicting effective thermal conductivity of the VHTR fuel block. However, in high temperature region above 500 .deg. C, the GAMMA+ model underestimates the effective thermal conductivity since radiation heat transfer is not reflected precisely. Further researches on it seem to be necessary.

  8. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  9. EUROCONTROL-Systemic Occurrence Analysis Methodology (SOAM)-A 'Reason'-based organisational methodology for analysing incidents and accidents

    Energy Technology Data Exchange (ETDEWEB)

    Licu, Tony [EUROCONTROL-European Organization for the Safety of Air Navigation and Dedale Asia Pacific, Safety Team, Rue de la Fusee, 96, 1130 Brussels (Belgium)]. E-mail: antonio.licu@eurocontrol.int; Cioran, Florin [EUROCONTROL-European Organization for the Safety of Air Navigation and Dedale Asia Pacific, Safety Team, Rue de la Fusee, 96, 1130 Brussels (Belgium)]. E-mail: florin.cioran@eurocontrol.int; Hayward, Brent [EUROCONTROL-European Organization for the Safety of Air Navigation and Dedale Asia Pacific, Safety Team, Rue de la Fusee, 96, 1130 Brussels (Belgium)]. E-mail: bhayward@dedale.net; Lowe, Andrew [EUROCONTROL-European Organization for the Safety of Air Navigation and Dedale Asia Pacific, Safety Team, Rue de la Fusee, 96, 1130 Brussels (Belgium)]. E-mail: alowe@dedale.net

    2007-09-15

    The Safety Occurrence Analysis Methodology (SOAM) developed for EUROCONTROL is an accident investigation methodology based on the Reason Model of organisational accidents. The purpose of a SOAM is to broaden the focus of an investigation from human involvement issues, also known as 'active failures of operational personnel' under Reason's original model, to include analysis of the latent conditions deeper within the organisation that set the context for the event. Such an approach is consistent with the tenets of Just Culture in which people are encouraged to provide full and open information about how incidents occurred, and are not penalised for errors. A truly systemic approach is not simply a means of transferring responsibility for a safety occurrence from front-line employees to senior managers. A consistent philosophy must be applied, where the investigation process seeks to correct deficiencies wherever they may be found, without attempting to apportion blame or liability.

  10. New developments in the German emergency planning as a consequence of Fukushima; Neue Entwicklungen im deutschen Notfallschutz nach Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Welte, Ulrike

    2015-06-01

    The analysis of the reactor accident in Fukushima Daiichi in March 2011 (lessons learned) in Germany and other countries has consequences for the national emergency planning in the respective governmental authorities. The contribution summarizes the most important aspects of the extended emergency plan concepts elaborated by the German SSK (Strahlenschutzkommission) and the RSK (Reaktorschutzkommission). The radiological principles were revised, recommendations concerning the monitoring and measuring equipment, accident scenarios for the emergency planning and measures for the post-accident phases are included in the concept.

  11. How does corruption influence perceptions of the risk of nuclear accidents?: cross-country analysis after the 2011 Fukushima disaster in Japan.

    OpenAIRE

    Yamamura, Eiji

    2011-01-01

    Japan’s 2011 natural disasters were accompanied by a devastating nuclear disaster in Fukushima. This paper used cross-country data obtained immediately after the Japanese disaster to explore how, and the extent to which, corruption affects the perception of citizens regarding the risk of nuclear accidents. Endogeneity bias was controlled for using instrumental variables. The cross-country analysis showed that citizens in less corrupt countries tend to perceive there to be a lower possibility ...

  12. Analysis of avalanche risk factors in backcountry terrain based on usage frequency and accident data in Switzerland

    Science.gov (United States)

    Techel, F.; Zweifel, B.; Winkler, K.

    2015-09-01

    Recreational activities in snow-covered mountainous terrain in the backcountry account for the vast majority of avalanche accidents. Studies analyzing avalanche risk mostly rely on accident statistics without considering exposure (or the elements at risk), i.e., how many, when and where people are recreating, as data on recreational activity in the winter mountains are scarce. To fill this gap, we explored volunteered geographic information on two social media mountaineering websites - bergportal.ch and camptocamp.org. Based on these data, we present a spatiotemporal pattern of winter backcountry touring activity in the Swiss Alps and compare this with accident statistics. Geographically, activity was concentrated in Alpine regions relatively close to the main Swiss population centers in the west and north. In contrast, accidents occurred equally often in the less-frequented inner-alpine regions. Weekends, weather and avalanche conditions influenced the number of recreationists, while the odds to be involved in a severe avalanche accident did not depend on weekends or weather conditions. However, the likelihood of being involved in an accident increased with increasing avalanche danger level, but also with a more unfavorable snowpack containing persistent weak layers (also referred to as an old snow problem). In fact, the most critical situation for backcountry recreationists and professionals occurred on days and in regions when both the avalanche danger was critical and when the snowpack contained persistent weak layers. The frequently occurring geographical pattern of a more unfavorable snowpack structure also explains the relatively high proportion of accidents in the less-frequented inner-alpine regions. These results have practical implications: avalanche forecasters should clearly communicate the avalanche danger and the avalanche problem to the backcountry user, particularly if persistent weak layers are of concern. Professionals and recreationists, on the

  13. Major accident analysis and prevention of coal mines in China from the year of 1949 to 2009

    Institute of Scientific and Technical Information of China (English)

    Wu Lirong; Jiang Zhongan; Cheng Weimin; Zuo Xiuwei; Lv Dawei; Yao Yujing

    2011-01-01

    From the year of 1949 to the present,the China national coal output has been increasing quickly and became first in the world in 2009.But at the same time,major coal mining accidents still exist nowadays.In order to review the overall situation and provide information on major accidents of coal mines in China,we investigated 26 major coal mining accidents in China between the years of 1949 and 2009 through statistical methods,each of which led to more than 1 00 fatalities.Statistical characteristics about accident-related factors such as time,death toll,accident reasons,characters and nature of enterprise were analyzed.And some special conclusions have been achieved.For example,although we have made great progress,the safety situation in China coal mining industry is still serious,and the reasons for the mining accidents are all human errors which are not inevitable.Such results may be helpful to prevent major accidents in coal mines.Moreso,based on both the knowledge of other countries which have good safety situation nowadays and the safety management situation of China,we made suggestion on safety management of China coal mining.In conclusion,countermeasures were proposed in accordance with the results of statistical studies and the analyses of problems existed in coal mines,including the perfection of safety supervision organization,the establishment of cooperating agency among government,coal mines and workers,the perfection of safety rules and regulations,the improvement of safety investment,the enhancement of safety training,the development of safety technique,and the development of emergency rescue technique and equipment.

  14. Labor Migration from Ukraine to the EU: an Analysis of Characteristics and Consequences

    Directory of Open Access Journals (Sweden)

    Bezrukova Nataliya V.

    2016-02-01

    Full Text Available The article is aimed at analyzing characteristics of labor migration from Ukraine to the European Union and evaluating the possible consequences of this process, as well as determining the future development trends in terms of emigration from Ukraine. The main directions of labor migration from Ukraine have been examined. An estimation of the volume of labor migration has been provided. Data about the socio-demographic characteristics of migrants have been presented. It is specified that the process of emigration has both positive and negative consequences for our country. Particular attention is paid to the analysis of employment settings of migrants in the Member States of the European Union. The main causes of emigration from Ukraine, among which both unemployment and low wages, have been allocated. The authors prove that the average monthly salary of Ukrainian migrants in the EU Member States is much higher than in Ukraine. On the basis of the carried out study has been concluded that labor migration from Ukraine to the EU has important socio-economic importance to our State. At the same time, signing of the Association agreement between Ukraine, on the one hand, and the European Union, the European Atomic Energy Community and their Member States, on the other hand, as well as introduction of a visa-free regime, will contribute to an increase in the number of labor migrants.

  15. Study on severe accident for traditional PWR based on RELAP5 and MELCOR combined analysis method%基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

    Institute of Scientific and Technical Information of China (English)

    王珏; 梁国兴

    2016-01-01

    针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用 RELAP5和 MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应.为了尽可能地利用 RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以 MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟.计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s.由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用 MELCOR 分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性.%A combined analysis method utilizing thermal-hydraulic system code RELAP5 and severe accident integral code MELCOR is developed to study the transient response of a traditional three-loop PWR under the severe accident TMLB’scenario. In order to utilize RELAP5 to the maximum degree and guarantee the accuracy of system response before entering into severe accident situation,the minimum cutoff temperature for zircaloy oxidation model of MELCOR,default value of 1 100 K,is used as the criterion to switch RELAP5 transient calculation to MELCOR severe accident analysis. Required data to initiate MELCOR will be extracted through the major edit of RELAP5 output. The results show that the data transferring process is relatively continuous. As observed in combined calculation,differences to varying degree are concluded

  16. Explosion risks and consequences for tunnels

    NARCIS (Netherlands)

    Weerheijm, J.; Berg, A.C. van den

    2014-01-01

    Tunnel accidents with transports of dangerous goods may lead to explosions. Risk assessment for these accidents is complicated because of the low probability and the unknown, but disastrous effects expected. Especially the lack of knowledge on the strength of the explosion and the consequences for t

  17. Fire and Thermal Effects of HD 1.3 Accidents: History, Research, and Analysis

    Science.gov (United States)

    2010-07-01

    5.3 Modeling Skin Burns 5.3.1 Model Description Skin consists of three layers: the epidermis , the dermis , and the subcutaneous tissue, see Figure...1, were used in the analysis. Table 1: Properties of skin tissues and blood Property Epidermis Dermis Subcutaneous Tissue Blood Thickness...that includes the effect of blood perfusion ( )bbb2 2 TTGρc x Tk t Tρc −− ∂ ∂ = ∂ ∂ Property Epidermis Dermis Subcutaneous Tissue Blood Thickness (m

  18. A cost-consequences analysis of a primary care librarian question and answering service.

    Directory of Open Access Journals (Sweden)

    Jessie McGowan

    Full Text Available BACKGROUND: Cost consequences analysis was completed from randomized controlled trial (RCT data for the Just-in-time (JIT librarian consultation service in primary care that ran from October 2005 to April 2006. The service was aimed at providing answers to clinical questions arising during the clinical encounter while the patient waits. Cost saving and cost avoidance were also analyzed. The data comes from eighty-eight primary care providers in the Ottawa area working in Family Health Networks (FHNs and Family Health Groups (FHGs. METHODS: We conducted a cost consequences analysis based on data from the JIT project. We also estimated the potential economic benefit of JIT librarian consultation service to the health care system. RESULTS: The results show that the cost per question for the JIT service was $38.20. The cost could be as low as $5.70 per question for a regular service. Nationally, if this service was implemented and if family physicians saw additional patients when the JIT service saved them time, up to 61,100 extra patients could be seen annually. A conservative estimate of the cost savings and cost avoidance per question for JIT was $11.55. CONCLUSIONS: The cost per question, if the librarian service was used at full capacity, is quite low. Financial savings to the health care system might exceed the cost of the service. Saving physician's time during their day could potentially lead to better access to family physicians by patients. Implementing a librarian consultation service can happen quickly as the time required to train professional librarians to do this service is short.

  19. The Effects of Consequence Manipulation during Functional Analysis of Problem Behavior Maintained by Negative Reinforcement

    Science.gov (United States)

    Potoczak, Kathryn; Carr, James E.; Michael, Jack

    2007-01-01

    Two distinct analytic methods have been used to identify the function of problem behavior. The antecedent-behavior-consequence (ABC) method (Iwata, Dorsey, Slifer, Bauman, & Richman, 1982/1994) includes the delivery of consequences for problem behavior. The AB method (Carr & Durand, 1985) does not include consequence delivery, instead relying…

  20. Research on Browsing Behavior in the Libraries: An Empirical Analysis of Consequences, Success and Influences

    Directory of Open Access Journals (Sweden)

    Shan-Ju L. Chang

    2000-12-01

    Full Text Available Browsing as an important part of human information behavior has been observed and investigated in the context of information seeking in the library in general and has assumed greater importance in human-machine interaction in particular. However, the nature and consequences of browsing are not well understood, and little is known of the success rate of such behavior.In this research, exploratory empirical case studies from three types of libraries were conducted, using questionnaires, observation logs, interviews, and computer search logs, to derive the empirical evidence to understand, from the user point of view, what are the consequences of browsing, what constitutes successful browsing, and what factors influence the extent of browsing. Content analysis and statistical analysis were conducted to analyze and synthesize the data. The research results show: (1 There are nine categories of the consequence of browsing, including accidental findings, modification of information need, found the desirable information, learning, feeling relaxation/recreational, information gathering, keeping updated, satisfying curiosity, and not finding what is needed. (2 Four factors that produce successful browsing: intention, the amount or quality of information, the utility of what is found, and help for solving problem or making judgment. (3 Three types of reasons for unsuccessful experience in browsing: not finding what one wanted, inadequate volume or quality of information, and not finding some things useful or interesting. (4 Three types of reasons for partial success: found the intended object but not happy with the quality or amount of information in it, not finding what one wanted but discovering new or potential useful information, not accomplish one purpose but achieve another one given multiple purposes. (5 The influential factors that affect the extent one engages in browsing include browser’s time, scheme of information organization, proximity to

  1. Choice & Consequence

    DEFF Research Database (Denmark)

    Khan, Azam

    between cause and effect in complex systems complicates decision making. To address this issue, we examine the central role that data-driven decision making could play in critical domains such as sustainability or medical treatment. We developed systems for exploratory data analysis and data visualization...... of data analysis and instructional interface design, to both simulation systems and decision support interfaces. We hope that projects such as these will help people to understand the link between their choices and the consequences of their decisions....

  2. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  3. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Ban, Chang Hwan; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Wang, Rongzhong; Yu, Hongxing [Nuclear Power Institute of China, Chengdu, SC (China)

    1994-12-01

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs.

  4. 高压加氢裂化装置事故模拟分析%Simulation Analysis on Accidents in High Pressure Hydrogenation Cracking Plant

    Institute of Scientific and Technical Information of China (English)

    张小红

    2012-01-01

    Taking the high pressure hydrogenation cracking plant as example,the plant was divided into 7 areas with potential hazard according to functions of each unit,and analysis was made on conditions of occurring fire or explosion accident in each area.Simulation was made for the accidents with SAFETI software,and corresponding safety precaution measures were raised based on data of accidents'affecting region.%以高压加氢裂化装置为例,根据各单元的功能,将装置划分为7个危险区域,对各区域发生火灾或爆炸事故的条件进行了分析。利用SAFETI软件对事故进行模拟,并根据事故影响范围数据提出了相应的安全措施。

  5. Causal Analysis of Mine Mechanical and Electrical Accidents and Countermeasures%基于煤矿机电事故原因探析及对策思考

    Institute of Scientific and Technical Information of China (English)

    郭凯

    2016-01-01

    在煤炭开采过程中由于人为因素和环境因素等多方面的影响,经常出现各种煤矿事故。本文主要对煤矿机电事故产生的原因进行分析,然后针对性提出一些策略,希望可以给煤矿相关研究人员提供参考。%Different kinds of coal mine accidents often occur due to construction method and influence of human factors and environmental factor,which not only influence the development of coal mine resources and threaten the life and health of mine workers.The paper mainly focuses on the causal analysis of mine mechanical and electrical accidents and presents some countermeasures.

  6. Squeal Those Tires! Automobile-Accident Reconstruction.

    Science.gov (United States)

    Caples, Linda Griffin

    1992-01-01

    Methods use to reconstruct traffic accidents provide settings for real life applications for students in precalculus, mathematical analysis, or trigonometry. Described is the investigation of an accident in conjunction with the local Highway Patrol Academy integrating physics, vector, and trigonometry. Class findings were compared with those of…

  7. Occupational blood exposure accidents in the Netherlands.

    NARCIS (Netherlands)

    Wijk, P.T.L. van; Schneeberger, P.M.; Heimeriks, K.; Boland, G.J.; Karagiannis, I.; Geraedts, J.; Ruijs, W.L.M.

    2010-01-01

    BACKGROUND: To make proper evaluation of prevention policies possible, data on the incidence and associated medical costs of occupational blood exposure accidents in the Netherlands are needed. METHODS: Descriptive analysis of blood exposure accidents and risk estimates for occupational groups. Cost

  8. The consequences of the Chernobyl accident in the Ukraine and problems with the sarcophagus; Die Folgen der Katastrophe von Tschernobyl in der Ukraine und die Probleme mit dem Sarkophag

    Energy Technology Data Exchange (ETDEWEB)

    Kopchinsky, G.A. [Atomaudit, Kiew (Ukraine)

    1996-07-01

    The reactor accident in the Ukraine contaminated part of the territory with iodine 131, caesium 137, strontium 90, and plutonium 239 and 240. The zone surrounding the site of the accident was declared restricted area; more than 90 000 persons were evacuated. The paper reports on current conditions in the restricted area and prospects for this area as well as on the current state of, and problems with, the sarcophagus. The conversion of the sarcophagus into an ecologically safe system and the economic situation of the Ukraine pose great problems. (DG) [Deutsch] Durch den Reaktorunfall in der Ukraine ist ein Teil des Territoriums mit Jod 131, Caesium 137, Strontium 90, Plutonium 239 und -240 kontaminiert worden. Um den Unfallort wurde eine Isolierungszone geschaffen und mehr als 90.000 Menschen evakuiert. Ueber den Zustand und die Perspektiven der Isolierungszone sowie ueber den Zustand und die Probleme des Sarkophags wird berichtet. Die Umgestaltung des Sarkophags in ein oekologisch sicheres System und die wirtschaftliche Situation der Ukraine bereiten grosse Probleme. (DG)

  9. 3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn Do Hee

    2003-01-01

    During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

  10. Emulation and Sobol' sensitivity analysis of an atmospheric dispersion model applied to the Fukushima nuclear accident

    Science.gov (United States)

    Girard, Sylvain; Mallet, Vivien; Korsakissok, Irène; Mathieu, Anne

    2016-04-01

    Simulations of the atmospheric dispersion of radionuclides involve large uncertainties originating from the limited knowledge of meteorological input data, composition, amount and timing of emissions, and some model parameters. The estimation of these uncertainties is an essential complement to modeling for decision making in case of an accidental release. We have studied the relative influence of a set of uncertain inputs on several outputs from the Eulerian model Polyphemus/Polair3D on the Fukushima case. We chose to use the variance-based sensitivity analysis method of Sobol'. This method requires a large number of model evaluations which was not achievable directly due to the high computational cost of Polyphemus/Polair3D. To circumvent this issue, we built a mathematical approximation of the model using Gaussian process emulation. We observed that aggregated outputs are mainly driven by the amount of emitted radionuclides, while local outputs are mostly sensitive to wind perturbations. The release height is notably influential, but only in the vicinity of the source. Finally, averaging either spatially or temporally tends to cancel out interactions between uncertain inputs.

  11. PF-4 simulated fire accident analysis: Filter-spray cool-down system reevaluation implications

    Energy Technology Data Exchange (ETDEWEB)

    White, B.W.; Gregory, W.S.

    1990-10-01

    The Los Alamos National Laboratory PF-4 facility was designed with spray cool down systems within the building's ventilation systems. The Engineering and Safety Analysis Group was asked, in cooperation with ENG-8 and MST-8, to evaluate whether the spray cool-down system still need to be classified as safety class'' systems. The study was performed using the FIRAC computer code. Given the fire source terms (hypothetical fire energy or time-temperature history), FIRAC can predict the pertinent transient flow parameters (pressures, flows, and temperatures) throughout a previously defined and selected fire zone. A computer model for the study that had all of the main ventilation systems in the south half of the PF-4 facility was used. Because the most hazardous room is located in the 400 Section, all ventilation systems but the 400 Section's one were simplified. The impetus for simplification was to keep the computer model tractable, and this was possible with the following assumptions: the fire cannot spread from one room to another, all corridor connecting doors are closed and will not fail under the pressures generated by the fire, and the principal pathway for potential release is the ventilation system. All of the blowers continue to operate, and all fire retardant systems fail to operate during the fire. The ASTM time-temperature curve was the source for the burn-room temperature, and smoke injection was used as input as well. Five different computer runs were made using different combinations of source terms and heat transfer. A connection from the burn room to the glovebox ventilation system was created by burning the glovebox plastic shielding; it was modeled by a branch having an initial flow 75 ft{sup 3}/min. 7 refs., 35 refs., 15 tabs.

  12. Consequences of bullying victimization in childhood and adolescence: A systematic review and meta-analysis

    Science.gov (United States)

    Moore, Sophie E; Norman, Rosana E; Suetani, Shuichi; Thomas, Hannah J; Sly, Peter D; Scott, James G

    2017-01-01

    AIM To identify health and psychosocial problems associated with bullying victimization and conduct a meta-analysis summarizing the causal evidence. METHODS A systematic review was conducted using PubMed, EMBASE, ERIC and PsycINFO electronic databases up to 28 February 2015. The study included published longitudinal and cross-sectional articles that examined health and psychosocial consequences of bullying victimization. All meta-analyses were based on quality-effects models. Evidence for causality was assessed using Bradford Hill criteria and the grading system developed by the World Cancer Research Fund. RESULTS Out of 317 articles assessed for eligibility, 165 satisfied the predetermined inclusion criteria for meta-analysis. Statistically significant associations were observed between bullying victimization and a wide range of adverse health and psychosocial problems. The evidence was strongest for causal associations between bullying victimization and mental health problems such as depression, anxiety, poor general health and suicidal ideation and behaviours. Probable causal associations existed between bullying victimization and tobacco and illicit drug use. CONCLUSION Strong evidence exists for a causal relationship between bullying victimization, mental health problems and substance use. Evidence also exists for associations between bullying victimization and other adverse health and psychosocial problems, however, there is insufficient evidence to conclude causality. The strong evidence that bullying victimization is causative of mental illness highlights the need for schools to implement effective interventions to address bullying behaviours.

  13. Accident knowledge and emergency management

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, B.; Groenberg, C.D.

    1997-03-01

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs.

  14. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

    1984-04-01

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

  15. Investigation of a fatal airplane crash: autopsy, computed tomography, and injury pattern analysis used to determine who was steering the plane at time of accident. A case report

    DEFF Research Database (Denmark)

    Høyer, Christian Bjerre; Nielsen, Trine Skov; Nagel, Lise Loft

    2012-01-01

    A fatal accident is reported in which a small single-engine light airplane crashed. The airplane carried two persons in the front seats, both of whom possessed valid pilot certificates. Both victims were subject to autopsy, including post-mortem computed tomography scanning (PMCT) prior to the au......A fatal accident is reported in which a small single-engine light airplane crashed. The airplane carried two persons in the front seats, both of whom possessed valid pilot certificates. Both victims were subject to autopsy, including post-mortem computed tomography scanning (PMCT) prior...... to the autopsy. The autopsies showed massive destruction to the bodies of the two victims but did not identify any signs of acute or chronic medical conditions that could explain loss of control of the airplane. PMCT, histological examination, and forensic chemical analysis also failed to identify an explanation...... for the crash. A detailed review of an airplane identical to the crashed airplane was performed in collaboration with the Danish Accident Investigation Board and the Danish National Police, National Centre of Forensic Services. The injuries were described using the abbreviated injury scale, the injury severity...

  16. Gas Station Oil Tank Explosion Accident Tree Analysis%加油站油罐区爆炸事故树分析

    Institute of Scientific and Technical Information of China (English)

    朱晋宇; 曹雄; 谷明朝; 韩苗苗; 王悦

    2013-01-01

      对油罐区爆炸的原因进行了详细的分类,绘制了油罐区爆炸事故树。通过结构重要度和最小割集分析得出,除达到爆炸极限外,油气积聚和油罐区通风不良是引发事故的重要因素,并提出防止油罐爆炸事故的有效可行方法评价的结果对加油站进行安全决策、减少事故隐患,具有积极的指导意义。%For oil tank area, the cause of the explosion carried on the detailed classification, on the basis of drawing the oil tank explosion ac⁃cident tree. Through the structure important degree, the minimum cut set analysis, reach the explosion limit, causing accumulation of oil and gas and oil tank area, poor ventilation is an important factor to cause an accident, and to prevent the oil tank explosion accident, effective and feasible method to improve the level of the oil tank area safety work, the results of the evaluation to the gas station for safety decision-making, reduce accidents, has a positive guiding significance.

  17. Corporate Cost of Occupational Accidents

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.; Impgaard, M.

    2004-01-01

    The systematic accident cost analysis (SACA) project was carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. Its focused on developing and testing a method for evaluating...

  18. Analysis of 129I in the soils of Fukushima Prefecture: preliminary reconstruction of 131I deposition related to the accident at Fukushima Daiichi Nuclear Power Plant (FDNPP).

    Science.gov (United States)

    Muramatsu, Yasuyuki; Matsuzaki, Hiroyuki; Toyama, Chiaki; Ohno, Takeshi

    2015-01-01

    Iodine-131 is one of the most critical radionuclides to be monitored after release from reactor accidents due to the tendency for this nuclide to accumulate in the human thyroid gland. However, there are not enough data related to the reactor accident in Fukushima, Japan to provide regional information on the deposition of this short-lived nuclide (half-life = 8.02 d). In this study we have focused on the long-lived iodine isotope, (129)I (half-life of 1.57 × 10(7) y), and analyzed it by accelerator mass spectrometry (AMS) for surface soil samples collected at various locations in Fukushima Prefecture. In order to obtain information on the (131)I/(129)I ratio released from the accident, we have determined (129)I concentrations in 82 soil samples in which (131)I concentrations were previously determined. There was a strong correlation (R(2) = 0.84) between the two nuclides, suggesting that the (131)I levels in soil samples following the accident can be estimated through the analysis of (129)I. We have also examined the possible influence from (129m)Te on (129)I, and found no significant effect. In order to construct a deposition map of (131)I, we determined the (129)I concentrations (Bq/kg) in 388 soil samples collected from different locations in Fukushima Prefecture and the deposition densities (Bq/m(2)) of (131)I were reconstructed from the results.

  19. [Analysis of knowledge attitudes and practices of health care workers facing blood exposure accidents in a general surgery service].

    Science.gov (United States)

    Ennigrou, Samir; Ben Ameur Khechine, Imène; Cherif, Ali; Najah, Nabil; Ben Hamida, Abdelmajid

    2004-06-01

    In order to assess the degree of knowledge, attitudes and the personnel's practices exercising in a service of general surgery of the hospital Charles Nicolle of Tunis, concerning blood exposure accidents, we did a transverse survey during the month of January of the year 2002. A questionnaire has been addressed to 114 people while using the technique of the direct interview. The middle age of investigated is 35.7 years. The sex ratio is 0.7. Only the 2/3 declare have been vaccinated against the B hepatitis. The results show a good knowledge of the exposure risk to a communicable disease by blood (95.6%), but less good for the risk of contamination by the three viruses HBV, HCV and HIV. The resheathing of needles, considered like gesture to risk, is underestimated by 71.2% of investigated. The majority of investigated declare to know universal precaution principles (85.8%). However, to the maximum 4 measures only on the 10 advisable have been mentioned by investigated. The conduct to hold in case of blood exposure accident seems insufficiently known by our sample. It is represented, in 78.8% of cases, in the application of disinfectants Betadine type or alcohol iodized, whereas the practice of a serology to the patient source is ignored completely. 75% of investigated having had a blood exposure accident lasting the last 12 months (n = 44) didn't declare their blood exposure accident and only 11.4% declare to have undergone cares. Actions of information and formation, to the intention of the whole of the personnel of the service, on risks incurred by the nursing, gestures and procedures to risk, the universal precaution respect, the conduct to hold in case of a blood exposure accident, the interest of the declaration and the interest of the vaccination against the B hepatitis, are primordial.

  20. Safety against releases in severe accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I.; Berg, Oe.; Nonboel, E. [eds.

    1997-12-01

    The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au) 39 refs.