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Sample records for pwr severe accident

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

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

    1992-03-01

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

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

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    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  3. Radiative heat transfer modelling in a PWR severe accident sequence

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    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    a debris bed. In particular, an expression of the conductivity was established in cells in which remaining cylinders and debris particles coexist. In the present document, after a recall of the main lines of the modelling, an application to a reactor sequence is proposed. A severe accident transient with core degradation is simulated. The radiative transfer model is shown to behave properly and to smoothly calculate the transitions between the successive core configurations. A comparison with the more classical Hottel method shows that the present model gives a better prediction of the degradation progression owing to a more accurate estimate of the radial heat transfers. References: [1] M. Zabiego et al., ICARE/CATHARE V1: application to a PWR 900 MWe severe accident sequence, SARJ, Tokyo, 1999; [2] M. Zabiego, F. Fichot, P. Rubiolo Transfert radiatif lors d'une sequence accidentelle dans un coeur de Reacteur a Eau sous Pression, Congres Francais de Thermique, SFT 2004, Presqu'ile de Giens, 25-28 mai 2004. (authors)

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

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    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.)

  5. Development of a parametric containment event tree model of a severe PWR accident

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    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  6. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

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    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

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

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    Johnson, S. C. [Westinghouse Electric Company LLC, 5801 Bluff Road, Columbia, SC 29209 (United States); Henry, R. E.; Paik, C. Y. [Fauske and Associates, Inc., 16W070 83rd Street, Burr Ridge, IL 60527 (United States)

    2012-07-01

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

  8. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

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    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  9. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

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    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  10. Characterization of PWR vessel steel tearing under severe accident condition temperatures

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    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  11. The study of core melting phenomena in reactor severe accident of PWR

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    Park, Jae Hong; Jeun, Gyoo Dong; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Bang, Kwang Hyun; Kim, Ki Yong [Korea Maritime Univ., Busan (Korea, Republic of)

    1999-03-15

    After TMI-2 accident, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining confidence in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression is proposed.

  12. The study of core melting phenomena in reactor severe accident of PWR

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    Park, Jae Hong; Jeun, Gyoo Dong; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Bang, Kwang Hyun; Kim, Ki Yong [Korea Maritime Univ., Busan (Korea, Republic of)

    1999-03-15

    After TMI-2 accident, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining confidence in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression is proposed.

  13. The study of core melting phenomena in reactor severe accident of PWR

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    Jeun, Gyoo Dong; Cho, Sung Won; Bang, Kwang Hyun; Park, Shane; Park, Seong Yong; Kim, Jin Man; Lim, Jae Hyuck; Song, Myung Jin [Hanyang Univ., Seoul (Korea, Republic of)

    2000-03-15

    TMI-2 accident is more valuable than the related experiments in the point of view that it is a real accident offering huge information about the late phase of severe accident. Therefore it gives out good standards for evaluation of code performance and inputs suitableness by comparing the accident data and simulated outputs. In this study SCDAP/REALAP5/MOD3.4 was selected for accident simulation. And sensitivity analysis was performed on varied cases to find out the most proper input variable about the late phase of core meting phenomena. Other plants and experimental facilities input deck were collected and analyzed for the sensitivity study and the shortcomings proposed by SCDAP/RELAP5 peer review were considered to the simulation. As a result gamma heating fraction in the input affect the progress of core melting phenomena. About this a study on the related model itself will be carried out.

  14. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

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    Mehboob, Khurram, E-mail: khurramhrbeu@gmail.com [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Xinrong, Cao [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ahmed, Raheel [College of Automation, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ali, Majid [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China)

    2013-09-15

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value.

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

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    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

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

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

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    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

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

  17. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

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    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  18. On-line measurement of gaseous iodine species during a PWR severe accident

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    Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

  19. On-line measurements of RuO{sub 4} during a PWR severe accident

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    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  20. Effect of water injection on hydrogen generation during severe accident in PWR

    Institute of Scientific and Technical Information of China (English)

    TAO Jun; CAO Xuewu

    2009-01-01

    Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.

  1. A study of core melting phenomena in reactor severe accident of PWR

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    Jeun, Gyoo Dong; Park, Shane; Kim, Jong Sun; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Jin Man [Korea Maritime Univ., Busan (Korea, Republic of)

    2001-03-15

    In the 4th year, SCDAP/RELAP5 best estimate input data obtained from the TMI-2 accident analysis were applied to the analysis of domestic nuclear power plant. Ulchin nuclear power plant unit 3, 4 were selected as reference plant and steam generator tube rupture, station blackout SCDAP/RELAP5 calculation were performed to verify the adequacy of the best estimate input parameters and the adequacy of related models. Also, System 80+ EVSE simulation was executed to study steam explosion phenomena in the reactor cavity and EVSE load test was performed on the simplified reactor cavity geometry using TRACER-II code.

  2. The reaction between iodine and organic coatings under severe PWR accident conditions. An experimental parameter study

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    Hellmann, S.; Funke, F.; Greger, G.U.; Bleier, A.; Morell, W. [Siemens AG, Power Generation Group, Erlangen (Germany)

    1996-12-01

    An extensive experimental parameter study was performed on the deposition and on the resuspension kinetics in the reaction system iodine/organically coated surfaces. Both reactions in the gas phase and in the liquid phase were investigated and kinetic rate constants suitable for modelling were derived. Previous experimental studies on the reaction of iodine with organic coated surfaces were mostly limited to temperatures below 100{sup o}C. Thus, this parameter study aims at filling a gap and providing kinetic data on heterogeneous reactions with organic surfaces in the accident-relevant temperature range of 100-160{sup o}C. Two types of laboratory experiments carried out at Siemens/KWU using coatings representative for German power plants (epoxy-tape paint), namely gas phase tests and liquid phase tests. (author) 6 figs., 6 tabs., 5 refs.

  3. Regulatory research of the PWR severe accident information needs and instrumentation availability for hydrogen control and management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae-Hong; Park, Gun-Chul; Suh, Kune Y.; Kang, Yun-Moon; Lee, Un-Jang; Oh, Se-Chul; Lee, Jin-Yong [Seoul Nationl Univ., Seoul (Korea, Republic of)

    1998-03-15

    During the current research period, we have set forth the methodology for identification of a severe accident, developed a framework for hydrogen management decision trees, and analyzed the literature on hydrogen management and experimental data for hydrogen bum. Specifically, we have summarized me results for information needs in a severe accident obtained in the U.S. and other countries, and applied the methodology to the reference plant YGN 3 and 4 as part of severe accident management. We have also examined the existing instruments in terms of their availability and survivability during a severe accident, and identified additionally needed information needs and instruments. We have identified dominant accident sequences for me reference plant YGN 3 and 4 to construct decision trees, and extracted available data from the IPE study of the plant. Based upon the data we have performed preliminary study on the decision tree and decision node. Last, we have examined various mechanisms for hydrogen generation and reIevant experimental data to predict me amount of hydrogen generation and governing factors in me process. We have also reviewed the hydrogen generation related models in the severe accident analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

    1993-09-01

    Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

  5. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kress, T. S. [comp.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  6. Study on Severe Accident Induced by Total Loss of Power Supply for Small PWR%小型压水堆完全丧失电源引发的严重事故研究

    Institute of Scientific and Technical Information of China (English)

    张龙飞; 舒礼伟; 陆古兵

    2012-01-01

    With the use of best estimate computer code RELAP/SCDAPS1M/MOD3. 4 of pressure water reactor severe accident, a three-channel along radial and ten-nodal along axis nuclear reactor severe accident calculation model was established based on a hypothetical small PWR. The severe accident induced by total loss of power supply was studied, and mitigation measure with 300 s continuation of the steam generator auxiliary feedwater was analyzed. The calculation results show that the steam generator auxiliary feedwater plays an important part in delaying core melt progression and mitigating severe accident consequences.%以压水堆严重事故最佳估算程序RELAP/SCDAPSIM/MOD3.4为核心软件,以假想的小型压水堆为研究对象,建立了1个径向3通道、轴向10节块的核反应堆严重事故计算模型,研究了完全丧失电源初因事件引发的严重事故过程,并对事故停堆后蒸汽发生器给水持续300 s的缓解措施进行了分析.计算结果表明:蒸汽发生器辅助给水对于延迟事故进程,缓解事故后果具有重要作用.

  7. Integral Test Facility PKL: Experimental PWR Accident Investigation

    OpenAIRE

    2012-01-01

    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  8. 压水堆严重事故后安全壳内辐射环境计算分析%Calculation and Analysis for the Radiation Condition in the Containment of PWR after Severe Accident

    Institute of Scientific and Technical Information of China (English)

    王晓霞; 张普忠; 刘新建

    2013-01-01

    In order to mitigate severe accident effectively,validation of equipment and instrument after severe accident need to be evaluated.The temperature,pressure,humidity and radiation are key parameters for the validation evaluation.For the source term released from the molten core to the containment,NUREG-1465 was adopted for PWR.Effect of spray and leakage on concentrations of radioactive nuclides in the containment was ignored.In this paper,γ and 3 radiation condition in the containment after severe accident were calculated and analyzed,which is very important to validation evaluation of equipment and instrument after severe accident.%为了确保有效的缓解严重事故,需要对用于缓解和监测严重事故进程的重要设备、仪表在严重事故环境下的可用性进行评估.而温度、压力、湿度、辐射等参数是可用性评估的重要输入条件.本文针对百万千瓦级压水堆核电机组,参考美国核管会发布的《轻水堆核电厂事故源项》(NUREG-1465)关于严重事故后放射性物质的释放阶段和释放份额的假设,计算出事故后由堆芯释放到安全壳内的放射性源项.对于放射性物质在安全壳内的分布,不考虑喷淋和泄漏的影响,计算并分析了严重事故后安全壳内的γ和β辐射环境条件,并与APl000的设备鉴定源项进行了对比分析.本文的计算对于设备和仪表在严重事故后的可用性分析以及其所需耐受的辐射条件具有重要的参考意义.

  9. Development of A Compact Severe Accident Simulator for PWR Nuclear Power Plants%压水堆核电站严重事故紧凑型仿真机开发

    Institute of Scientific and Technical Information of China (English)

    唐钢; 张森如; 江光明; 傅霄华

    2001-01-01

    为了缓解压水堆核电站可能发生的严重事故的后果,也为了满足安全分析工程师和概率风险评价人员的需求,并在与国际原子能机构合作框架协议内,研制开发了紧凑型的严重事故仿真分析机 MELSIM-PC。该仿真系统主要由仿真核心程序、同步通讯程序、人机界面程序等几个部分组成,可以工作在一台普通的微型计算机上,成功地实现 MELCOR程序变量的运行数据库管理、电站动态图形显示、仿真计算控制、再启动和仿真重演等重要功能。%In order to alleviate the consequence of a possible severe accident in PWR Nuclear Power Plants and in response to the demands of safety analysis engineers and Probabilistic Safety Assessment(PSA) specialists,a compact severe accident simulator has been developed under an IAEA TC project.The PC-based simulator consists of the database engine MELCOR code,the man-machine interface modules MANAGER & DISPLAY,the communication module SERVER and the supplementary modules.It can be used successfully to realize some very important functions,such as the variable database management of MELCOR code,the plant mimic screens,simulation computation control,restart and replay,etc.

  10. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  11. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    1989-11-01

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

  12. 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

  13. 核电厂严重事故下卸压对氢气产生的影响分析%Effect of Depressurization on Hydrogen Generation During Severe Accident in PWR Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    陶俊; 李京喜; 佟立丽; 曹学武

    2011-01-01

    研究了1 000 MWe压水堆核电厂在典型的高压严重事故序列下卸压对氢气产生的影响.分析结果表明,开启1列、2列和3列卸压阀进行一回路卸压均会在堆芯熔化进程的3个阶段导致氢气产生率的明显增大:1)堆芯温度1 500~2 100 K;2)堆芯温度2 500~2 800 K;3)从形成由硬壳包容的熔融池(2 800 K)到熔融物向压力容器下封头下落.开启卸压阀的列数越多,氢气产生率的增大越明显.%The effect of depressurization on hydrogen generation during a typical high pressure severe accident sequence in a 1 000 MWe pressurized water reactor (PWR) nuclear power plant was analyzed. Analyses results indicate that the hydrogen generation rate is obviously increased by the reactor coolant system depressurization of opening one, two or three power operated relief valves (PORVs) at three core damage states.The first is peak core temperature from 1 500 K to 2 100 K. The second is peak core temperature from 2 500 K to 2 800 K. The third is from formation of molten pool supported by crust to slumping of molten materials into reactor pressure vessel lower head.The more PORVs are opened the more increment of hydrogen generation rate.

  14. Water Reflooding Effectiveness Assessment for 1 000 MWe PWR under Severe Accident Condition%百万千瓦级压水堆严重事故后再注水的有效性评价

    Institute of Scientific and Technical Information of China (English)

    胡啸; 黄挺; 裴杰; 陈炼

    2015-01-01

    根据现有的设计资料,使用一体化严重事故分析程序 MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08 cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10 kg/s)、中流量(50 kg/s)和大流量(200 kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10 kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200 kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。%The MELCOR1.8.6 code was applied to a severe accident model of a 1 000 MWe PWR which includes primary system,secondary system,passive core cool-ing system and containment system.For the transient case,a small break LOCA with 2 inch (5.08 cm)break at the cold leg concurrent with failure of gravity injection was selected.After the core was damaged due to the failure of gravity inj ection,it was assumed that the coolant was inj ected into the pressure vessel,and then the water reflooding effectiveness was evaluated and analyzed.In this calculation,the coolant injection into reactor core with the small (10 kg/s),medium (50 kg/s)and large (200 kg/s)mass flow rates respectively at 3 different time stages of the severe accident was simulated.The effectiveness of water reflooding was assessed through hydrogen production,radioactive materials released from core,and core temperature.The results show that the mass flow rate above 10 kg/s is believed to be efficient for cooling a 1 000 MWe reactor at the beginning of core damage.However,with the accident devel-oping to core relocation,a large mass flow

  15. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  16. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  17. Iodine behaviour in severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Dutton, L.M.C.; Grindon, E.; Handy, B.J.; Sutherland, L. [NNC Ltd., Knutsford (United Kingdom); Bruns, W.G.; Sims, H.E. [AEA Technology, Harwell (United Kingdom); Dickinson, S. [AEA Technology, Winfrith (United Kingdom); Hueber, C.; Jacquemain, D. [IPSN/CEA, Cadarache, Saint Paul-Lez-Durance (France)

    1996-12-01

    A description is given of analyses which identify which aspects of the modelling and data are most important in evaluating the release of radioactive iodine to the environment following a potential severe accident at a PWR and which identify the major uncertainties which affect that release. Three iodine codes are used namely INSPECT, IODE and IMPAIR, and their predictions are compared with those of the PSA code MAAP. INSPECT is a mechanistic code which models iodine behaviour in the aqueous aerosol, spray water and sump water, and the partitioning of volatile species between the aqueous phases and containment gas space. Organic iodine is not modelled. IODE and IMPAIR are semi-empirical codes which do not model iodine behaviour in the aqueous aerosol, but model organic iodine. The fault sequences addressed are based on analyses for the Sizewell `B` design. Two types of sequence have been analysed.: (a) those in which a major release of fission products from the primary circuit to the containment occur, e.g. a large LOCAS, (b) those where the release by-passes the containment, e.g. a leak into the auxiliary building. In the analysis of the LOCA sequences where the pH of the sump is controlled to be a value of 8 or greater, all three codes predict that the oxidation of iodine to produce gas phase species does not make a significant contribution to the source term due to leakage from the reactor building and that the latter is dominated by iodide in the aerosol. In the case where the pH of the sump is not controlled, it is found that the proportion of gas phase iodine increases significantly, although the cumulative leakage predicted by all three codes is not significantly different from that predicted by MAAP. The radiolytic production of nitric acid could be a major factor in determining the pH, and if the pH were reduced, the codes predict an increase in gas phase iodine species leaked from the containment. (author) 4 figs., 7 tabs., 13 refs.

  18. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    OpenAIRE

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  19. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Energy Technology Data Exchange (ETDEWEB)

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  20. 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)

  1. 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.

  2. Porosity effects during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Cazares R, R. I. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Posgrado en Energia y Medio Ambiente, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Espinosa P, G.; Vazquez R, A., E-mail: ricardo-cazares@hotmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

  3. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  4. IVR-ERVC effectiveness assessment for large size advanced PWR under severe accident%严重事故下大功率先进压水堆IVR-ERVC有效性分析

    Institute of Scientific and Technical Information of China (English)

    金越; 刘晓晶; 程旭; 陈薇

    2016-01-01

    通过压力容器外部冷却(ERVC)以实现堆内熔融物滞留(IVR)作为反应堆严重事故缓解管理的一项重要举措一直以来广泛受到关注和研究.本文使用严重事故分析程序 MELCOR,从瞬态角度对大型先进压水堆进行了 IVR-ERVC相关研究.过程中重点关注了堆芯熔毁和重新定位,熔池形成、生长及其传热过程,并且对压力容器外部流动传热进行了分析.MELCOR计算所得下封头热流密度分布的瞬态结果与临界热流密度(CHF)比较和分析表明,1700 MWe 大功率压水堆发生严重事故后在 IVR-ERVC条件下能够保证压力容器的完整性,即,IVR-ERVC 能够有效带出下封头熔融物的衰变热量,缓解严重事故后果.%As a key severe accident management strategy for light water reactors (LWRs),in-vessel retention (IVR)through external reactor vessel cooling (ERVC)has been the focus of relevant studies for decades. This paper addressed the IVR-ERVC issues from a transient perspective using the severe accident code MELCOR for large size advanced passive power plant. Current analysis was mainly focused on the transients in severe accident including core degradation and relocation,molten pool formation,growth and heat transfer within,together with external flow and heat transfer analysis. MELCOR calculations for lower head heat flux were then compared with critical heat flux (CHF)of lower head to assess the effectiveness of IVR-ERVC. The results suggest that lower head heat flux is well below the CHF value. Thus,the IVR-ERVC strategy is considered to be physically effective.

  5. Severe accident simulation at Olkiuoto

    Energy Technology Data Exchange (ETDEWEB)

    Tirkkonen, H.; Saarenpaeae, T. [Teollisuuden Voima Oy (TVO), Olkiluoto (Finland); Cliff Po, L.C. [Micro-Simulation Technology, Montville, NJ (United States)

    1995-09-01

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

  6. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  7. ASTEC V2.0 reactor applications on French PWR 900 MWe accident sequences and comparison with MAAP4

    Energy Technology Data Exchange (ETDEWEB)

    Lombard, Virginie; Azarian, Garo; Ducousso, Erik; Gandrille, Pascal, E-mail: pascal.gandrille@areva.com

    2014-06-01

    In the frame of the SARNET Severe Accident Network of Excellence an important task of partners is the assessment of the ASTEC integral code, considered today as the European reference code for evaluation of the source term. A code-to-code comparison between ASTEC V2.0 rev1 and MAAP 4.0.7 code versions has been performed by AREVA NP SAS on a French PWR 900 MWe. Two transients have been analyzed, focussing on in-vessel phenomena: total loss of feedwater (H2 sequence in the French nomenclature) and total loss of onsite and offsite power (H3 sequence). The detailed analysis shows an overall good agreement between both code results on thermal-hydraulics, hydrogen production and core degradation phenomena.

  8. 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.

  9. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

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

  11. 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...

  12. 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.

  13. Severe Accident Recriticality Analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

    1999-11-01

    Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

  14. The Possibility of Building Nuclear Power Plant Free from Severe Accident Risk PWR NPP with advanced all passive safety cooling systems (AAP SCS)%发展无严重事故风险核电站的曙光具有完全非能动安全冷却系统的压水堆核电站

    Institute of Scientific and Technical Information of China (English)

    肖宏才

    2013-01-01

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP,actuated by natural force has been put forward in the article.Here the natural force mainly means the fore,which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another,including occurrence of accident situation.Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident,so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink.There is no need to rely on automatic control system,any active equipment and human actions in all working process of the AAP SCS,which can reduce the probability of severe accident to zero,so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety.Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology.So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk,and for modernization of existing second generation nuclear power plant.%本文提出了用自然力直接触发启动压水堆核电站一整套完全非能动的停堆安全冷却系统.这里的自然力主要是指一回路运行工况转换时由于其压力分布变化所形成的压差力.在这一系统中,当进行停堆或发生某种一回路事故工况时,相应的安全冷却系统便自然地投入运行,立即缓解事故后果,将事故时一回路释放的能量及堆芯余热非能动地排入最终热阱.在全过程中不依靠自动控制系统、能动设备及任何人为因素的介入,即可确保对堆芯余热无限期的安全冷却能力,完全避免压水堆核电站发生向环境泄漏放射性物

  15. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  16. A Bayesian ensemble of sensitivity measures for severe accident modeling

    Energy Technology Data Exchange (ETDEWEB)

    Hoseyni, Seyed Mohsen [Department of Basic Sciences, East Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Vagnoli, Matteo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge, Fondation EDF – Electricite de France Ecole Centrale, Paris, and Supelec, Paris (France); Pourgol-Mohammad, Mohammad [Department of Mechanical Engineering, Sahand University of Technology, Tabriz (Iran, Islamic Republic of)

    2015-12-15

    Highlights: • We propose a sensitivity analysis (SA) method based on a Bayesian updating scheme. • The Bayesian updating schemes adjourns an ensemble of sensitivity measures. • Bootstrap replicates of a severe accident code output are fed to the Bayesian scheme. • The MELCOR code simulates the fission products release of LOFT LP-FP-2 experiment. • Results are compared with those of traditional SA methods. - Abstract: In this work, a sensitivity analysis framework is presented to identify the relevant input variables of a severe accident code, based on an incremental Bayesian ensemble updating method. The proposed methodology entails: (i) the propagation of the uncertainty in the input variables through the severe accident code; (ii) the collection of bootstrap replicates of the input and output of limited number of simulations for building a set of finite mixture models (FMMs) for approximating the probability density function (pdf) of the severe accident code output of the replicates; (iii) for each FMM, the calculation of an ensemble of sensitivity measures (i.e., input saliency, Hellinger distance and Kullback–Leibler divergence) and the updating when a new piece of evidence arrives, by a Bayesian scheme, based on the Bradley–Terry model for ranking the most relevant input model variables. An application is given with respect to a limited number of simulations of a MELCOR severe accident model describing the fission products release in the LP-FP-2 experiment of the loss of fluid test (LOFT) facility, which is a scaled-down facility of a pressurized water reactor (PWR).

  17. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  18. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

    Frid, W.; Højerup, C.F.; Lindholm, I.

    2001-01-01

    three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto I plant in Finland...... with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power...... generation-for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s(-1) injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high...

  19. 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

  20. 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.

  1. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident; Developpement d`un element fini coque. Application au comportement thermo-viscoplastique d`une cuve de reacteur nucleaire (REP) en situation d`accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, V

    1998-10-07

    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author) 113 refs.

  2. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  3. 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)

  4. 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

  5. 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.

  6. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

  7. Interactions of severe accident research and regulatory positions (ISARRP)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. (comp.) [Royal Inst. of Tech., Stockholm (Sweden). Nuclear Power Safety

    2001-12-01

    in assessment of plant safety. This work package was also designed to distinguish the differences between the attitudes and approaches followed by the various regulatory organisations in Europe, Eastern Europe, USA and Japan. Work Package 5: Relevance of example PSA results to SA research. The objective of their work package was to employ the results of some recent PSAs (preferably for a PWR and a BWR) and relate their findings to the results obtained in SA research, and to the effectiveness of the SAM measures already taken or contemplated. Work Package 6: The state of resolution of the SA issues with respect to the needs. The objective of this work package is to have another look at the state of the resolution of the severe accident issues which have been identified over the years, and relate that to what the needs of the regulatory organizations are in terms of their functions. Work Package 7: Regulatory use of the results of severe accident research. The objective is to identify the results of the SA research which the regulatory organizations, over the years, have used in either defining specific regulatory actions or in not taking specific actions. Work Package 8: Remaining issues and concerns. The objective of the work here is to review the work in the previous work package and identify what are the remaining unresolved safety issues and concerns for which sufficient results of the SA research are not available. Work Package 9: Recommendations on future directions of severe accident research. The purpose of this work package is to provide recommendations to E.U. (and to the readers) by the authors of this report on the directions that should be followed, in the future for the conduct of severe accident research. These recommendations are in essence the conclusions of this study.

  8. 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.

  9. 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

  10. [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.

  11. simulation of a SGTR severe PWR-W with MELCOR code; Simulacion de un SGTR severo en un PWR-W con el codigo Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, A. J.; Jimenez Varas, G.; Israelsson, L. C.

    2014-04-01

    Steam Generator tube Rupture (SGTR) is a small break loss of coolant accident. the issues related to this kind of transients makes them different from the classics LOCA studies. SGTR accidents in Pressurized Water Reactor are known to be one of the most demanding transients for the operating crew. It this accident is not managed in a proper way it could lead to steam generator overfill and a severe accident inside containment . To simulate this accident the MELCOR code was chosen, whose aim is the assessment of the progression of severe accidents in Light Water Reactors. (Author)

  12. Severe Accident Test Station Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL; Terrani, Kurt A [ORNL

    2015-06-01

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

  13. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    Energy Technology Data Exchange (ETDEWEB)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment.

  14. ANS severe accident program overview & planning document

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.

    1995-09-01

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

  15. Simulation of a SGTR severe PWR-W with the MELCOR code; Simulacion de un SGTR severo en un PWR-W con el codigo MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, A. J.; Israelsson, C.; Jimenez, G.

    2013-07-01

    The type SGTR accident is a case of loss of coolant accident small features which make it necessary to differentiate and evolution of classical studies LOCA sequence type. To simulate this type of accident has chosen the MELCOR code, which aims to study the progression of severe accidents in LWR plants. It has been developed by Sandia National Laboratories for the United States Nuclear Regulatory Commission.

  16. Nuclear power plant Severe Accident Research Plan

    Energy Technology Data Exchange (ETDEWEB)

    Larkins, J T; Cunningham, M A

    1983-01-01

    The Severe Accident Research Plan (SARP) will provide technical information necessary to support regulatory decisions in the severe accident area for existing or planned nuclear power plants, and covers research for the time period of January 1982 through January 1986. SARP will develop generic bases to determine how safe the plants are and where and how their level of safety ought to be improved. The analysis to address these issues will be performed using improved probabilistic risk assessment methodology, as benchmarked to more exact data and analysis. There are thirteen program elements in the plan and the work is phased in two parts, with the first phase being completed in early 1984, at which time an assessment will be made whether or not any major changes will be recommended to the Commission for operating plants to handle severe accidents. Additionally at this time, all of the thirteen program elements in Chapter 5 will be reviewed and assessed in terms of how much additional work is necessary and where major impacts in probabilistic risk assessment might be achieved. Confirmatory research will be carried out in phase II to provide additional assurance on the appropriateness of phase I decisions. Most of this work will be concluded by early 1986.

  17. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

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

  18. 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

  19. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  20. 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

  1. Ranking of severe accident research priorities

    Energy Technology Data Exchange (ETDEWEB)

    Schwinges, B. [Gesell Anlagen and Reaktorsicherheit GRS mbH, D-50667 Cologne (Germany); Journeau, C. [CEA Cadarache, DEN STRI LMA, F-13115 St Paul Les Durance (France); Haste, T. [Paul Scherrer Inst, NES LTH, OVGA 312, CH-5232 Villigen (Switzerland); Meyer, L.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Trambauer, K. [GRS mbH, Forschungsgelande, D-85748 Garching (Germany)

    2010-07-01

    The objectives of the SARNET network are to define common research programmes in the field of severe accidents and to develop common computer tools and methodologies for safety assessment in this field. To reach these objectives, one of the work packages, named 'Severe Accident Research Priorities' (SARP), aimed at reviewing and reassessing the priorities of research issues as a basis to harmonize and to re-orient research programmes, to define new ones, and to close - if possible - resolved issues on a common basis. The work was performed in close collaboration with 8 participating institutions, led by GRS, representing technical safety organisations, industry and utilities (IRSN, CEA, EDF, FZK, GRS, KTH, TUS, VTT). This action made use notably of (1) the outcomes of the EURSAFE project in the 5. Framework Programme, i. e. the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the validation and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other SARNET tasks, and (4) the outcome of the research performed in the three thematic sub-domains of SARNET (corium, containment and source term). The main outcome of EURSAFE was a list of 21 topics which included recommendations for experimental programmes and code developments. This list formed the basis of the work in SARP. Also the methodology applied in EURSAFE to consider both the risk potential and the severe accident issues where large uncertainties still subsist was adopted. The analyses of the progress of research and development activities considered whether (1) any research issue was resolved due to reduction of uncertainties or gain of scientific insights, (2) any new issue had to be added to the list of needed research, (3) any new process or phenomenon had to be included in the general PIRT list taking into account the safety relevance and the lack of knowledge, and (4) any new accident management program has to be

  2. 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.

  3. Calculation of absorbed doses to water pools in severe accident sequences

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F. [Oak Ridge National Lab., TN (United States)

    1991-12-01

    A methodology is presented for calculating the radiation dose to a water pool from the decay of uniformly distributed nuclides in that pool. Motivated by the need to accurately model radiolysis reactions of iodine, direct application is made to fission product sources dissolved or suspended in containment sumps or pools during a severe nuclear reactor accident. Two methods of calculating gamma absorption are discussed - one based on point-kernal integration and the other based on Monte Carlo techniques. Using least-squares minimization, the computed results are used to obtain a correlation that relates absorbed dose to source energy and surface-to-volume ratio of the pool. This correlation is applied to most relevant fission product nuclides and used to actually calculate transient sump dose rate in a pressurized-water reactor (PWR) severe accident sequence.

  4. 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

  5. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  6. Severe Accidents in the Energy Sector

    Energy Technology Data Exchange (ETDEWEB)

    Hirschberg, S.; Spiekerman, G.; Dones, R

    1998-11-01

    A comprehensive database on severe accidents, with main emphasis on the ones associated with the energy sector, has been established by the Paul Scherrer Institute (PSI). Fossil energy carriers, nuclear power and hydro power are covered in ENSAD (Energy related Severe Accident Database), and the scope of work includes all stages of the analysed energy chains, i.e. exploration, extraction, transports, processing, storage and waste disposal. The database has been developed using a wide variety of sources. As opposed to the previous studies the ambition of the present work has been, whenever feasible, to cover a relatively broad spectrum of damage categories of interest. This includes apart from fatalities also serious injuries, evacuations, land or water contamination, and economic losses. Currently, ENSAD covers 13,914 accidents, of which 4290 are energy related, and 1943 are considered as severe accidents. Significant effort has been directed towards the examination of the relevance of the worldwide accident records to the Swiss specific conditions, particularly in the context of nuclear and hydro power. For example, a detailed investigation of large dam failures and their consequences was carried out. Generally, while Swiss specific aspects are emphasised, the major part of the collected and analysed data, as well as the insights gained, are considered to be of general interest. In particular, three sets of the aggregated results are provided based on world wide occurrence, on OECD countries, and on non OECD countries, respectively. Significant differences exist between the aggregated, normalised damage rates assessed for the various energy carriers: On the world wide basis, the broader picture obtained by coverage of full energy chains leads to aggregated immediate fatality rates being much higher for the fossil fuels than what one would expect if power plants only were considered. The highest rates apply to LPG, followed by hydro, oil, coal, natural gas and

  7. Analysis of the containment of a compact reactor PWR submitted to loss of coolant accident; Analise da contencao de um reator PWR compacto submetido a acidente de perda de refrigerante

    Energy Technology Data Exchange (ETDEWEB)

    Dutra, Alexandre de Souza; Belchior Junior, Antonio; Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    In the present paper analyses were done with the computer code RELAP5/MOD2 for rising the process conditions of the containment of a compact reactor PWR of low potency, submitted to Loss of Coolant Accidents (LOCA). The main results obtained were the behavior of maximum conditions of pressure as a function of the available containment free volume. It was also studied the problem of containment sub-compartmentation, that is to say, the possibility of the rupture to happen in restricted spaces generating high sub-compartment peak pressure and, consequently, high strains on the internal structures. (author)

  8. 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.

  9. SARNET: Severe accident research network of excellence

    Energy Technology Data Exchange (ETDEWEB)

    Albiol, T.; Van Dorsselaere, J. P. [IRSN, DPAM, F-13115 St Paul Les Durance (France); Chaumont, B. [IRSN, DSR, SAGR, F-92262 Fontenay Aux Roses (France); Haste, T. [Paul Scherrer Inst, NES, LTH, OVGA 312, CH-5232 Villigen (Switzerland); Journeau, Ch. [CEA Cadarache, DEN, STRI, LMA, F-13115 St Paul Les Durance (France); Meyer, L. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Sehgal, Bal Raj [KTH, AlbaNova Univ Ctr, S-10691 Stockholm (Sweden); Schwinges, Bernd [Gesell Anlagen and Reaktorsicherheit GRS mbH, D-50667 Cologne (Germany); Beraha, D. [GRS mbH, Forschungsgelande, D-85748 Garching (Germany); Annunziato, A. [Commiss European Communities, JRC, IPSC, I-21020 Ispra, VA (Italy); Zeyen, R. [Commiss European Communities, JRC IE, IRSN DPAM DIR, F-13115 St Paul Les Durance (France)

    2010-07-01

    Fifty-one organisations network in SARNET (Severe Accident Research Network of Excellence) their research capacities in order to resolve the most important pending issues for enhancing, with regard to Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). This project. co-funded by the European Commission (EC) under the 6. Framework Programme, has been defined in order to optimise the use of the available means and to constitute sustainable research groups in the European Union. SARNET tackles the fragmentation that may exist between the different national R and D programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the organisations involved in SA research in Europe, plus Canada. To reach these objectives, all the organisations networked in SARNET contributed to a joint Programme of Activities, which consisted of: Implementation of an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; Harmonization and re-orientation of the research programmes, and definition of new ones; Analysis of the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; Development of the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; Development of Scientific Databases in which all the results of research programmes are stored in a common format (DATANET); Development of a common methodology for Probabilistic Safety Assessment of NPPs; Development of short courses and writing a textbook on Severe Accidents for students and researchers; Promotion of personnel mobility amongst various European organisations. This paper presents the major achievements after four and a half years of operation of the

  10. ACR-1000 design provisions for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  11. Quantification of severe accident source terms of a Westinghouse 3-loop plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee Min [Department of Engineering and System Science, and Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec II, Kung Fu Road, Hsinchu, Taiwan (China)], E-mail: mlee@mail.ess.nthu.edu.tw; Ko, Y.-C. [Department of Engineering and System Science, and Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec II, Kung Fu Road, Hsinchu, Taiwan, ROC (China)

    2008-04-15

    Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment.

  12. 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

  13. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-15

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  14. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  15. 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)

  16. Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, Joy L. [Rempe and Associates, Idaho Falls, ID (United States); Knudson, Darrell L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

  17. Applicability of simplified human reliability analysis methods for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2016-03-15

    Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

  18. Development of severe accident management and training support system

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Sub; Kim, Ko Ryo; Jung, Won Dae; Ha, Jae Joo

    2001-04-01

    Recently, the overall severe accident management strategy is under development according to the logical flow of severe accident management guidelines in some foreign countries. In Korea, the basis of severe accident management strategy is established due to the development of Korean severe accident guideline. In the straining system, the professional information as well as the general information for severe accident should be provided to the related personnel and the function of prior simulation for plant behavior according to strategy execution should be required. Korean severe accident management guideline is chosen as the basis logic for development of support system for decision-support and training related with execution of severe accident strategy. The training simulator is developed for prior expectation of plant behavior and the severe accident computer code, MELCOR, is utilized as the engine, and it is possible to operate equipments necessary for execution of severe accident management guidelines. And also, the graphical interface is developed to provide the plant status and provide status change of major equipments dynamically.

  19. On severe accident hydrogen behaviour in Loviisa

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-02-01

    This study is related to the hydrogen management strategy of the Loviisa ice-condenser containments. A synthetic survey is conducted of the various parts of the subject by using compact `back-of-the-envelope` analysis methods. The analysed cases are consistent with the principal hydrogen management approaches proposed by the utility Imatran Voima Oy (IVO). The study begins by introduction of the Loviisa plant features and various severe accident types. Hydrogen generation characteristics are analysed mainly for the core degradation phase, but the hydrogen sources from molten fuel-coolant interactions and reflooding of a degraded core are discussed, as well. The hydrogen generation and release rates are compared with the overall gas convection and mixing conditions in order to estimate hydrogen concentrations in the containment. The natural convection currents are examined also from the scaling point of view, concerning the scaled-down VICTORIA tests of IVO. Finally, the potential for large deflagration loadings or local detonations is examined for the Loviisa containments. The study is concluded by preliminary subjective judgments about the most critical factors of the Loviisa hydrogen problematics and about any issues that may require additional confirmative research. (orig.) (47 refs., 4 figs., 24 tabs.).

  20. Iodine chemical forms in LWR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Beahm, E.C.; Weber, C.F.; Kress, T.S.; Parker, G.W.

    1991-01-01

    Calculated data from seven severe accident sequences in light-water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I{sup {minus}}). The radiation induced conversion of I{sup {minus}} in water pools into I{sub 2} is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled, and large fractions of iodine as I{sub 2} within the containment atmosphere may be produced. 16 refs.

  1. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of strategies there may be several options available to the operator; and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrument behavior during an accident. During the period September 26--28, 1990, a workshop was held at the University of California, Los Angeles, to address these uncertainties for Boiling Water Reactors (BWRs). This report contains a summary of the workshop proceedings.

  3. MELCOR simulation of postulated severe accidents in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seongn Yeon; Kim Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hwan Yeol; Park, Jong Hwa [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Since the Fukushima accident in 2011, severe accidents of a nuclear power plant have been a target of big debate whether the defense in depth philosophy applied to current nuclear system is still vigorous enough to ensure the protection of the operators and the public. Thus an accurate prediction of severe accident has become a critical task for the nuclear engineers with reliable employment of Probabilistic Risk Analysis (PRA). According to a recent PRA result, Small Break Loss Of Coolant Accident (SBLOCA) without safety injection and Station Black Out (SBO) show high probability of proceeding to severe accidents. Thus, these accident scenarios need to be evaluated properly with reliable prediction tools. Song and Ahn analyzed SBO sequences in KSNP using MELCOR 1.8.5. Park and Song examined SBLOCA scenarios based on the PSA of KNSP using MAAP 4.06. Their studies utilized severe accident database. In continuation of the further analysis, several scenarios of postulated SBO and SBLOCA in OPR1000 are investigated using the severe accident database and MELCOR 1.8.6.

  4. Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading

    Directory of Open Access Journals (Sweden)

    J. Arndt

    2012-01-01

    Full Text Available During postulated high-pressure core melt accident scenarios, temperature values of more than 800°C can be reached in the reactor coolant line and the surge line of a pressurised water reactor (PWR, before the bottom of the reactor pressure vessel experiences a significant temperature increase due to core melting. For the assessment of components of the primary cooling circuit, two methods are used by GRS. One is the simplified method ASTOR (approximated structural time of rupture. This method employs the hypothesis of linear damage accumulation for modeling damage progression. A failure time surface which is generated by structural finite element (FE analysis of varying pressure and temperature loads serves as a basis for estimations of failure times. The second method is to perform thermohydraulic and structure mechanic calculations for the accident scenario under consideration using complex calculation models. The paper shortly describes both assessment procedures. Validation of the ASTOR method concerning a large-scale test on a pipe section with geometric properties similar to a reactor coolant line is presented as well as severe accident scenarios investigated with both methods.

  5. Predicting Severity and Duration of Road Traffic Accident

    Directory of Open Access Journals (Sweden)

    Fang Zong

    2013-01-01

    Full Text Available This paper presents a model system to predict severity and duration of traffic accidents by employing Ordered Probit model and Hazard model, respectively. The models are estimated using traffic accident data collected in Jilin province, China, in 2010. With the developed models, three severity indicators, namely, number of fatalities, number of injuries, and property damage, as well as accident duration, are predicted, and the important influences of related variables are identified. The results indicate that the goodness-of-fit of Ordered Probit model is higher than that of SVC model in severity modeling. In addition, accident severity is proven to be an important determinant of duration; that is, more fatalities and injuries in the accident lead to longer duration. Study results can be applied to predictions of accident severity and duration, which are two essential steps in accident management process. By recognizing those key influences, this study also provides suggestive results for government to take effective measures to reduce accident impacts and improve traffic safety.

  6. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

  7. 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).

  8. Prediction of the Containment Pressure under Severe Accidents Using CFNN

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Young Do; Choi, Geon Pil; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    It is important to keep containment integrity by measuring main risk factors such as temperature and hydrogen concentration that occur pressure rise in the containment and by operating safety features at the right time. In this study, the circumstance that instrumentation equipment in NPPs is uncertain under severe accidents after DBA is assumed. This is to keep containment integrity by manually generating the safety injection actuation signal (SIAS) and to assess integrity of accident equipment through early prediction of the containment pressure under extreme circumstances when main factors such as temperature and hydrogen concentration that rise pressure in containment may not be adequately measured. In this study, the cascaded fuzzy neural network (CFNN) model is used to predict containment pressure using LOCA break sizes as input data. Because the real severe accident data cannot be obtained from actual NPP accidents, they were gained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using modular accident analysis program (MAAP) code. Temperature and hydrogen concentration in the containment are risk factors that increase containment pressure under severe accidents. Therefore, LOCA sizes as the input data for the CFNN model are used to predict the containment pressure. This input data is the simulation data obtained by using MAAP4 code for the OPR1000 reactor. As a result of using the CFNN model, the RMS errors are within 0.4% to 1.4. Accordingly, The CFNN could be a model that reliably predict the containment pressure and the data through CFNN model could figure out the containment integrity and assess the survivability of severe accident equipment under accidents.

  9. 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

  10. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

  11. Source term analyses under severe accidents for KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    2001-03-01

    In this study, in-containment source term for LOFW (Loss of Feed Water), which has appeared the most frequent core melt accident, is calculated and compared with NUREG-1465 source term. This study provides not only new source term data using MELCOR1.8.4 and its state-of-the-art models but also evaluating basis of KNGR design and its mitigation capability under severe accidents. As the selected accident is identical with LOFW-S17, which has been analyzed using MAAP by KEPCO with only difference of 2 SITs, mutual comparison of the results is especially expected.

  12. Severities of transportation accidents involving large packages

    Energy Technology Data Exchange (ETDEWEB)

    Dennis, A.W.; Foley, J.T. Jr.; Hartman, W.F.; Larson, D.W.

    1978-05-01

    The study was undertaken to define in a quantitative nonjudgmental technical manner the abnormal environments to which a large package (total weight over 2 tons) would be subjected as the result of a transportation accident. Because of this package weight, air shipment was not considered as a normal transportation mode and was not included in the study. The abnormal transportation environments for shipment by motor carrier and train were determined and quantified. In all cases the package was assumed to be transported on an open flat-bed truck or an open flat-bed railcar. In an earlier study, SLA-74-0001, the small-package environments were investigated. A third transportation study, related to the abnormal environment involving waterways transportation, is now under way at Sandia Laboratories and should complete the description of abnormal transportation environments. Five abnormal environments were defined and investigated, i.e., fire, impact, crush, immersion, and puncture. The primary interest of the study was directed toward the type of large package used to transport radioactive materials; however, the findings are not limited to this type of package but can be applied to a much larger class of material shipping containers.

  13. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) 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 has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  14. 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)

  15. Drug use and the severity of a traffic accident

    NARCIS (Netherlands)

    Smink, BE; Ruiter, B; Lusthof, KJ; de Gier, JJ; Uges, DRA; Egberts, ACG

    Several studies have showed that driving under the influence of alcohol and/or certain illicit or medicinal drugs increases the risk of a (severe) crash. Data with respect to the question whether this also leads to a more severe accident are sparse. This study examines the relationship between the

  16. Drug use and the severity of a traffic accident

    NARCIS (Netherlands)

    Smink, BE; Ruiter, B; Lusthof, KJ; de Gier, JJ; Uges, DRA; Egberts, ACG

    2005-01-01

    Several studies have showed that driving under the influence of alcohol and/or certain illicit or medicinal drugs increases the risk of a (severe) crash. Data with respect to the question whether this also leads to a more severe accident are sparse. This study examines the relationship between the u

  17. Radiological Consequence Analyses Following a Hypothetical Severe Accident in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Juyub; Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In order to reflect the lessons learned from the Fukushima Daiichi nuclear power plant accident, a simulator which is named NANAS (Northeast Asia Nuclear Accident Simulator) for overseas nuclear accident has been developed. It is composed of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. For the source-term estimation module, the representative reactor types were selected as CPR1000, BWR5 and BWR6 for China, Japan and Taiwan, respectively. Considering the design characteristics of each reactor type, the source-term estimation module simulates the transient of design basis accident and severe accident. The atmospheric dispersion prediction module analyzes the transport and dispersion of radioactive materials and prints out the air and ground concentration. Using the concentration result, the dose assessment module calculates effective dose and thyroid dose in the Korean Peninsula region. In this study, a hypothetical severe accident in Japan was simulated to demonstrate the function of NANAS. As a result, the radiological consequence to Korea was estimated from the accident. PC-based nuclear accident simulator, NANAS, has been developed. NANAS contains three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. The source-term estimation module simulates a nuclear accident for the representative reactor types in China, Japan and Taiwan. Since the maximum calculation speed is 16 times than real time, it is possible to estimate the source-term release swiftly in case of the emergency. The atmospheric dispersion prediction module analyzes the transport and dispersion of radioactive materials in wide range including the Northeast Asia. Final results of the dose assessment module are a map projection and time chart of effective dose and thyroid dose. A hypothetical accident in Japan was simulated by NANAS. The radioactive materials were released during the first 24 hours and the source

  18. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  19. TRAC analyses of severe overcooling transients for the Oconee-1 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ireland, J R [comp.

    1985-05-01

    This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

  20. Contribution of the Exposure Pathways After a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Hwang, Wontae; Han, Moonhee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    A radiological dose assessment calculates the amount of radiation energy absorbed by a potentially exposed individual as a result of a specific exposure. Public can be exposure from several exposure pathways. External doses occur when the body is exposed to radioactive material outside the body. When making the emergency preparedness for severe accident from NPPs, therefore, we need to have comprehension about those exposure pathways. Thus, in this study, an evaluation of external and internal dose from radioactive materials during severe accident was performed to find out exposure pathway from which the dose has the highest value for several radionuclides. The basic study to make out the relation between exposure pathways and dose from them was performed. In the emergency phase, the most affecting nuclide type on public was noble gas, especially {sup 133}Xe, and the dominant exposure pathway was could shine. Also, in the long term-phase, the most affecting nuclide type on public was fission product, especially {sup 90}Sr, and the dominant exposure pathway was water ingestion. The information of the dose composition from exposure pathway obtained in this study might be basic data for making emergency preparedness plan for severe accident. In the future, assessment of the source term is expected to enhance the reliability of dose assessment during severe accident.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  2. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  3. 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)

  4. Severe accident natural circulation studies at the INEL

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, P.D.; Brownson, D.A.; Dobbe, C.A.; Jones, K.R.; O`Brien, J.E.; Pafford, D.J.; Schlenker, L.D.; Tung, V.X.

    1995-02-01

    Severe accident natural circulation flows have been investigated at the Idaho National Engineering Laboratory to better understand these flows and their potential impacts on the progression of a pressurized water reactor severe accident. Parameters affecting natural circulation in the reactor vessel and hot legs were identified and ranked based on their perceived importance. Reviews of the scaling of the 1/7-scale experiments performed by Westinghouse were undertaken. RELAP5/MOD3 calculations of two of the experiments showed generally good agreement between the calculated and observed behavior. Analyses of hydrogen behavior in the reactor vessel showed that hydrogen stratification is not likely to occur, and that an initially stratified layer of hydrogen would quickly mix with a recirculating steam flow. An analysis of the upper plenum behavior in the Three Mile Island, Unit 2 reactor concluded that vapor temperatures could have been significantly higher than the temperatures seen by the control rod drive lead screws, supporting the premise that a strong natural circulation flow was likely present during the accident. SCDAP/RELAP5 calculations of a commercial pressurized water reactor severe accident without operator actions showed that the natural circulation flows enhance the likelihood of ex-vessel piping failures long before failure of the reactor vessel lower head.

  5. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

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

  6. Experiments on silver-indium-cadmium control rod failure during severe accident sequences

    Energy Technology Data Exchange (ETDEWEB)

    Steinbrueck, M.; Stegmaier, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2010-05-15

    Silver-indium-cadmium (SIC) alloy is used as neutron absorber material in control rods (CR) of Pressurized Water Reactors (PWR). It is the material with the lowest melting temperature (approx. 1100 K) among all metallic and ceramic materials applied in nuclear reactors. During a hypothetical severe accident the SIC melt is kept in its stainless steel (SS) cladding tube as long as this is intact. After failure of the cladding tube by eutectic interaction with the Zircaloy-4 (Zry-4) guide tube or latest by reaching the SS melting temperature SIC elements are released and may interact with other core components. Furthermore, Ag-In-Cd are one of the main contributors to aerosol release in the reactor cooling system and may strongly influence nature and transport of fission products in the primary circuit and later on in the containment. The bundle experiment QUENCH-13 with prototypical SIC control rod as well as two series of single-rod tests with 10-cm long CR segments were performed at Karlsruhe Institute of Technology (KIT, former FZK) in order to improve the data base on SIC CR degradation and aerosol release. This paper concentrates on the degradation and failure mechanisms of SIC CRs as well as on the interaction between SIC absorber melt with other core components. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  8. The reaction between iodine and silver under severe PWR accident conditions. An experimental parameter study

    Energy Technology Data Exchange (ETDEWEB)

    Funke, F.; Greger, G.U.; Bleier, A.; Hellmann, S.; Morell, W. [Siemens AG, Power Generation Group, Erlangen (Germany)

    1996-12-01

    An extensive experimental parameter study was performed on the kinetics in the reaction system I{sub 2}/Ag and I{sup -}/Ag in a laboratory-scale apparatus.Starting with I{sub 2} or I{sup -} solutions and silver powder suspensions, the decrease of soluted I{sub 2} or I{sup -}, respectively, due to fixation on the silver particles, was monitored as function of time using the radioactive tracer I-131. The measured data were analyzed using a model of first order kinetics with respect to the iodine concentration. However, the analysis using first order kinetics had to be performed separately in an early, fast reaction phase and in a late, slow reaction phase. The reason for this unexpected behaviour was not identified. Thus, rate constant, two for each test, were deduced from 14 I{sub 2}/Ag main tests and from 36 I{sup -}/Ag tests. No dependencies of the rate constants were found on the parameters temperature, initial iodine concentration, presence of boric acid, type of silver educt, and pretreatment of the silver educt prior to the tests. However, the stirring of the reaction solution generally enhanced the kinetics highlighting the importance of mass transfer. The I{sup -}/Ag reaction proceeded only if there was no inertization of the reaction solution by sparging with nitrogen. The temperature-independent rate constant for the early, fast I{sub 2}/Ag reaction phase is 2E-5 m/s. However, a smaller rate constant of 6E-6 m/s is recommended for use in source term calculations with IMPAIR, which already contains a first order model. Analogously, the temperature-independent I{sup -}/Ag reaction rate constant is 8E-6 m/s in an early, fast reaction phase. For use in source term calculations, a smaller rate constant of 2E-6 m/s is recommended. The lower bound of the I{sup -}/Ag rate constant was 3E-8 m/s which could be used in very conservative source term calculations. (author) 20 figs., 6 tabs., 15 refs.

  9. On the performance of an artificial bee colony optimization algorithm applied to the accident diagnosis in a PWR nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali S. de; Schirru, Roberto; Medeiros, Jose A.C.C., E-mail: maghali@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b, E-mail: canedo@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2009-07-01

    The swarm-based algorithm described in this paper is a new search algorithm capable of locating good solutions efficiently and within a reasonable running time. The work presents a population-based search algorithm that mimics the food foraging behavior of honey bee swarms and can be regarded as belonging to the category of intelligent optimization tools. In its basic version, the algorithm performs a kind of random search combined with neighborhood search and can be used for solving multi-dimensional numeric problems. Following a description of the algorithm, this paper presents a new event classification system based exclusively on the ability of the algorithm to find the best centroid positions that correctly identifies an accident in a PWR nuclear power plant, thus maximizing the number of correct classification of transients. The simulation results show that the performance of the proposed algorithm is comparable to other population-based algorithms when applied to the same problem, with the advantage of employing fewer control parameters. (author)

  10. Influence of radiation heat transfer during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Cazares R, R. I.; Epinosa P, G.; Varela H, J. R.; Vazquez R, A. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Polo L, M. A., E-mail: ricardo-cazares@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    The aim of this work is to determine the influence of the radiation heat transfer on an average fuel channel during a severe accident of a BWR nuclear power plant. The analysis considers the radiation heat transfer in a participating medium, where the gases inside the system participate in the radiation heat transfer. We consider the steam-water mixture as an isothermal gray gas, and the boundaries of the system as a gray diffuse isothermal surface for the clad and refractory surfaces for the rest, and consider the average fuel channel as an enclosure system. During a severe accident, generation and diffusion of hydrogen begin at high temperature range (1,273 to 2,100 K), and the fuel rod cladding oxidation, but the hydrogen generated do not participate in the radiation heat transfer because it does not have any radiation properties. The heat transfer process in the fuel assembly is considered with a reduced order model, and from this, the convection and the radiation heat transfer is introduced in the system. In this paper, a system with and without the radiation heat transfer term was calculated and analyzed in order to obtain the influence of the radiation heat transfer on the average fuel channel. We show the behavior of radiation heat transfer effects on the temporal evolution of the hydrogen concentration and temperature profiles in a fuel assembly, where a stream of steam is flowing. Finally, this study is a practical complement for more accurate modeling of a severe accident analysis. (Author)

  11. Steam Oxidation Testing in the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.

  12. 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.

  13. Radiological environment within an NPP after a severe nuclear accident

    Science.gov (United States)

    Andgren, Karin; Fritioff, Karin; Buhr, Anna Maria Blixt; Huutoniemi, Tommi

    2017-09-01

    The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room) can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level). Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere). Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

  14. Predictions of structural integrity of steam generator tubes under normal operating, accident, an severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States)

    1997-02-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation was confirmed by further tests at high temperatures, as well as by finite-element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation was confirmed by finite-element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate-sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure was developed and validated by tests under various temperature and pressure loadings that can occur during postulated severe accidents.

  15. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    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. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  16. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  17. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  18. Analysis of flammability in the attached buildings to containment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

    2016-11-15

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H{sub 2} and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  19. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  20. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  1. 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.

  2. Severe head injury caused by motorcycle traffic accident

    Institute of Scientific and Technical Information of China (English)

    李钢

    1999-01-01

    Objective To explore the characteristic and treatment of the severe head injury due to motorcycle accident.Methods Review and analysis of 27 motorcycle traffic trauma cases who were admitted to our hospital from Oct.1995 to Sep.1997.Results Young men were the main composition of these patients.Multiple injuries associated with brain ste or diffuse axonal injury were common,which were the main factors influencing the consciousness and prognosis of the patients.The wound was usually severely contaminated.Evacuation of hematomas,decompression by depleting skull flap,hypotheymia and artificial hibernation were conducted in this series.Among them,14 cases were cured ,3 cases were seriously disabled,10 cases died.Conclusions Motorcycle's weight is light so it easily loses its balance.The riders and the passengers are exposed and lack protection.Driving against traffic regulations is frquently seen.All these are the reasons why the motorcycle traffic accidents often take place. When the traffic accident happens,the patients' head generally is thrown a long distance and dashed against the barrier or the ground.The psture nd mechanism of injury were complicated and varied.The decelerated injury and rolling injury occurred frequently and they were the main reasons for brain stem or diffuse axonal injury.The patients who have surgical indication should be operated upon as soon as possible.Hibernation and low temoerature therapy are conducive to the protection of the brain function at the early stage of postinjury or postoperation.A careful epluchage is essential to reduce infection of the open injury.

  3. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

    A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

  4. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, H. D.; Park, S. Y.; Kim, K. R.; Park, S. H.; Choi, Y.; Song, Y. M.; Ahn, K. I.; Park, J. H

    2005-04-01

    The scope of the project includes four activities such as construction of DB, development of data base management tool, development of severe accident analysis code system and FP studies. In the construction of DB, level-1,2 PSA results and plant damage states event trees were mainly used to select the following target initiators based on frequencies: LLOCA, MLOCA, SLOCA, station black out, LOOP, LOFW and SGTR. These scenarios occupy more than 95% of the total frequencies of the core damage sequences at KSNP. In the development of data base management tool, SARD 2.0 was developed under the PC microsoft windows environment using the visual basic 6.0 language. In the development of severe accident analysis code system, MIDAS 1.0 was developed with new features of FORTRAN-90 which makes it possible to allocate the storage dynamically and to use the user-defined data type, leading to an efficient memory treatment and an easy understanding. Also for user's convenience, the input (IEDIT) and output (IPLOT) processors were developed and implemented into the MIDAS code. For the model development of MIDAS concerning the FP behavior, the one dimensional thermophoresis model was developed and it gave much improvement to predict the amount of FP deposited on the SG U-tube. Also the source term analysis methodology was set up and applied to the KSNP and APR1400.

  5. An Evaluation Methodology Development and Application Process for Severe Accident Safety Issue Resolution

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2012-01-01

    Full Text Available A general evaluation methodology development and application process (EMDAP paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management guidelines. The basic framework described in this paper extends the top-down, bottom-up strategy described in the U.S Nuclear Regulatory Commission Regulatory Guide 1.203 to severe accident evaluations addressing U.S. NRC expectation for plant design certification applications.

  6. 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.

  7. Use of decision trees for evaluating severe accident management strategies in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

    2016-07-15

    Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

  8. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    Directory of Open Access Journals (Sweden)

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  9. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  10. 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.

  11. 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.

  12. Development of ultrasonic high temperature system for severe accidents research

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Young Ro and others

    2000-07-01

    The aims of this study are to find a gap formation between corium melt and the reactor lower head vessel, to verify the principle of the gap formation and to analyze the effect of the gap formation on the thermal behavior of corium melt and the lower plenum. This report aims at suggesting development of a new high temperature measuring system using an ultrasonic method which overcomes the limitations of the present thermocouple method used for severe accident experiments. Also, this report describes the design and manufacturing method of the ultrasonic system. At that time, the sensor element is fabricated to a reflective element using 1mm diameter and 50 mm and 80 mm long tungsten alloy wires. This temperature measuring system is intended to measure up to 2800 deg C.

  13. Influence diagrams and decision trees for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, W.W.J.

    1996-09-01

    A review of relevant methodologies based on Influence Diagrams (IDs), Decision Trees (DTs), and Containment Event Trees (CETs) was conducted to assess the practicality of these methods for the selection of effective strategies for Severe Accident Management (SAM). The review included an evaluation of some software packages for these methods. The emphasis was on possible pitfalls of using IDs and on practical aspects, the latter by performance of a case study that was based on an existing Level 2 Probabilistic Safety Assessment (PSA). The study showed that the use of a combined ID/DT model has advantages over CET models, in particular when conservatisms in the Level 2 PSA have been identified and replaced by fair assessments of the uncertainties involved. It is recommended to use ID/DT models complementary to CET models. (orig.).

  14. Bus accident severity and passenger injury: evidence from Denmark

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    Purpose Bus safety is a concern not only in developing countries, but also in the U.S. and Europe. In Denmark, disentangling risk factors that are positively or negatively related to bus accident severity and injury occurrence to bus passengers can contribute to promote safety as an essential...... picture of the bus safety situation in Denmark and suggest the necessity of further research into bus drivers’ attitudes and perceptions of risks and road users’ perceptions of bus operations. Moreover, these findings sug- gest the need for further training into bus drivers’ hazard recognition skills...... light. Occurrence of injury to bus passengers is positively related to (i) the involvement of heavy vehicles, (ii) crossing intersections in yellow or red light, (iii) open areas, (iv) high speed limits, and (v) slippery road surface. Conclusions The findings of the current study provide a comprehensive...

  15. Factors associated with the severity of construction accidents: The case of South Australia

    Directory of Open Access Journals (Sweden)

    Jantanee Dumrak

    2013-12-01

    Full Text Available While the causes of accidents in the construction industry have been extensively studied, severity remains an understudied area. In order to provide more evidence for the currently limited number of empirical investigations on severity, this study analysed 24,764 construction accidents reported during 2002-11 in South Australia. A conceptual model developed through literature uses personal characteristics such as age, experience, gender and language. It also employs work-related factors such as size of organization, project size and location, mechanism of accident and body location of the injury. These were shown to discriminate why some accidents result in only a minor severity while others are fatal. Factors such as time of accident, day of the week and season were not strongly associated with accident severity. When the factors affecting severity of an accident are well understood, preventive measures could be developed specifically to those factors that are at high risk.

  16. A study on the late core melt progression in pressurized water reactor severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Jeun Gyoo Dong; Bang, Kwang Hyun; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Back, Hyung Hmm [Korea Maritime Univ., Busan (Korea, Republic of)

    1998-03-15

    After TMI-2 accidents, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression os proposed.

  17. Development of the severe accident risk information database management system SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies.

  18. Severe accident research and management in Nordic Countries - A status report

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, SKI (Sweden)] (ed.)

    2002-01-01

    The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

  19. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  20. Project on Transfer Mechanism of Radioactive Source Term Under Severe Accident

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

    <正>The "Transfer mechanism of radioactive source term under severe accident" is a sub-project of the research program of "Mechanism and phenomenology of severe accident". An aerosol transfer mechanism experimental facility is built to simulate the passive containment cooling system (PCCS) of advanced pressurizer reactors to research effects to the transfer process of fission products under severe accident. An advanced CFD method is also utilized to research the effects. The objective of this project is to understand

  1. A database system for the management of severe accident risk information, SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K. I.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies.

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

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

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

  3. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  4. 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.

  5. 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.

  6. Spatial Analysis of Accident Spots Using Weighted Severity Index ...

    African Journals Online (AJOL)

    ADOWIE PERE

    Density-based Clustering for Traffic Accident Risk (DBCTAR) was carried out to assist in ascertaining the distribution of ... least one road vehicle, occurring on a road open to ... Road Safety Agency (FRSC), the Lagos State Traffic. Management ...

  7. Prediction of hydrogen concentration in containment during severe accidents using fuzzy neural network

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Kim, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-03-15

    Recently, severe accidents in nuclear power plants (NPPs) have become a global concern. The aim of this paper is to predict the hydrogen buildup within containment resulting from severe accidents. The prediction was based on NPPs of an optimized power reactor 1,000. The increase in the hydrogen concentration in severe accidents is one of the major factors that threaten the integrity of the containment. A method using a fuzzy neural network (FNN) was applied to predict the hydrogen concentration in the containment. The FNN model was developed and verified based on simulation data acquired by simulating MAAP4 code for optimized power reactor 1,000. The FNN model is expected to assist operators to prevent a hydrogen explosion in severe accident situations and manage the accident properly because they are able to predict the changes in the trend of hydrogen concentration at the beginning of real accidents by using the developed FNN model.

  8. Proceedings of the workshop on severe accident research held in Japan (SARJ-97)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1998-05-01

    The Workshop on Severe Accident Research held in Japan (SARJ-97) was taken place at Pacifico Yokohama on October 6 - 8, 1997, and attended by 180 participants from 15 countries and one international organizations. The 59 papers, which cover wide areas of severe accident research both in experiments and analysis, such as in-vessel melt retention, fuel-coolant interaction, fission products behavior, structural integrity, containment behavior, computer simulations, and accident management, are indexed individually. (J.P.N.)

  9. Proceedings of the workshop on severe accident research held in Japan (SARJ-98)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1999-07-01

    The Workshop on Severe Accident Research held in Japan (SARJ-98) was taken place at Hotel Lungwood on November 4-6, 1998, and attended by 181 participants from 13 countries. The 63 papers, which cover wide areas of severe accident research both in experiments and analyses, such as in-vessel melt retention, fuel-coolant interaction, fission products behavior, structural integrity, containment behavior, computer simulations, and accident management, are indexed individually. (J.P.N.)

  10. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    Energy Technology Data Exchange (ETDEWEB)

    Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  11. 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.

  12. Development of a prototype graphic simulation program for severe accident training

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo

    2000-05-01

    This is a report of the development process and related technologies of severe accident graphic simulators, required in industrial severe accident management and training. Here, we say 'a severe accident graphic simulator' as a graphics add-in system to existing calculation codes, which can show the severe accident phenomena dynamically on computer screens and therefore which can supplement one of main defects of existing calculation codes. With graphic simulators it is fairly easy to see the total behavior of nuclear power plants, where it was very difficult to see only from partial variable numerical information. Moreover, the fast processing and control feature of a graphic simulator can give some opportunities of predicting the severe accident advancement among several possibilities, to one who is not an expert. Utilizing graphic simulators' we expect operators' and TSC members' physical phenomena understanding enhancement from the realistic dynamic behavior of plants. We also expect that severe accident training course can gain better training effects using graphic simulator's control functions and predicting capabilities, and therefore we expect that graphic simulators will be effective decision-aids tools both in sever accident training course and in real severe accident situations. With these in mind, we have developed a prototype graphic simulator having surveyed related technologies, and from this development experiences we have inspected the possibility to build a severe accident graphic simulator. The prototype graphic simulator is developed under IBM PC WinNT environments and is suited to Uljin 3and4 nuclear power plant. When supplied with adequate severe accident scenario as an input, the prototype can provide graphical simulations of plant safety systems' dynamic behaviors. The prototype is composed of several different modules, which are phenomena display module, MELCOR data interface module and graphic database

  13. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... COMMISSION Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors... comment on NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power..., ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' DATES: Submit comments...

  14. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... COMMISSION Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors... the Commission), issued a NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports...), Section 19.0 ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' The NRC...

  15. Statistical modelling of the frequency and severity of road accidents

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    of the reasons for heterogeneity has been made, which in the end may lead to devising policy measures (Paper 1). 3) A connection between the occurrence probability of trauma type and crash, vehicle and person characteristics exists (Paper 2). 4) The attitudes that accident reporting is useless are found...... the literature about under-reporting and gives new and innovative knowledge which can contribute to new policy measures for improving the reporting rate. For that reason this thesis should be used as an important tool whenever addressing the under-reporting challenge....... management tool.Initially models were built by using existing traffic accident data collected by the police and emergency rooms in Denmark. The data registered by the police was collected on traffic accidents occurred on Danish roads in the period between 2002 and 2008. The emergency room data were collected...

  16. 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.

  17. The special severity of occupational accidents in the afternoon: "the lunch effect".

    Science.gov (United States)

    Camino López, Miguel A; Fontaneda, Ignacio; González Alcántara, Oscar J; Ritzel, Dale O

    2011-05-01

    The severity of occupational accidents suffered by construction workers at different hours of the day is analyzed in this study. It may be seen that the interval of time between 13:00 and 17:00 has incomprehensibly high rates of severe and fatal accidents in comparison with any other. We associate this higher accident rate with what we have termed the "lunch effect". We studied 10,239,303 labor accidents in Spain over the period 1990-2002. The relationships between potential risk factors for occupational accidents around lunch in Spain, especially alcohol consumption are studied, using two methods: analysis of national archival data of 2,155,954 occupational accidents suffered by workers in the construction sector over the period 1990-2002 and a survey study. This study also seeks to contribute the opinions of the workers themselves regarding the causes that might explain this situation.

  18. Radiation protection issues on preparedness and response for a severe nuclear accident: experiences of the Fukushima accident.

    Science.gov (United States)

    Homma, T; Takahara, S; Kimura, M; Kinase, S

    2015-06-01

    Radiation protection issues on preparedness and response for a severe nuclear accident are discussed in this paper based on the experiences following the accident at Fukushima Daiichi nuclear power plant. The criteria for use in nuclear emergencies in the Japanese emergency preparedness guide were based on the recommendations of International Commission of Radiological Protection (ICRP) Publications 60 and 63. Although the decision-making process for implementing protective actions relied heavily on computer-based predictive models prior to the accident, urgent protective actions, such as evacuation and sheltering, were implemented effectively based on the plant conditions. As there were no recommendations and criteria for long-term protective actions in the emergency preparedness guide, the recommendations of ICRP Publications 103, 109, and 111 were taken into consideration in determining the temporary relocation of inhabitants of heavily contaminated areas. These recommendations were very useful in deciding the emergency protective actions to take in the early stages of the Fukushima accident. However, some suggestions have been made for improving emergency preparedness and response in the early stages of a severe nuclear accident.

  19. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  20. Study of the Severity of Accidents in Tehran Using Statistical Modeling and Data Mining Techniques

    Directory of Open Access Journals (Sweden)

    Hesamaldin Razi

    2013-01-01

    Full Text Available AbstractBackgrounds and Aims: The Tehran province was subject to the second highest incidence of fatalities due to traffic accidents in 1390. Most studies in this field examine rural traffic accidents, but this study is based on the use of logit models and artificial neural networks to evaluate the factors that affect the severity of accidents within the city of Tehran.Materials and Methods: Among the various types of crashes, head-on collisions are specified as the most serious type, which is investigated in this study with the use of Tehran’s accident data. In the modeling process, the severity of the accident is the dependent variable and defined as a binary covariate, which are non-injury accidents and injury accidents. The independent variables are parameters such as the characteristics of the driver, time of the accident, traffic and environmental characteristics. In addition to the prediction accuracy comparison of the two models, the elasticity of the logit model is compared with a sensitivity analysis of the neural network.Results: The results show that the proposed model provides a good estimate of an accident's severity. The explanatory variables that have been determined to be significant in the final models are the driver’s gender, age and education, along with negligence of the traffic rules, inappropriate acceleration, deviation to the left, type of vehicle, pavement conditions, time of the crash and street width.Conclusion: An artificial neural network model can be useful as a statistical model in the analysis of factors that affect the severity of accidents. According to the results, human errors and illiteracy of drivers increase the severity of crashes, and therefore, educating drivers is the main strategy that will reduce accident severity in Iran. Special attention should be given to a driver’s age group, with particular care taken when they are very young.

  1. A statistical description of the types and severities of accidents involving tractor semi-trailers

    Energy Technology Data Exchange (ETDEWEB)

    Clauss, D.B.; Wilson, R.K. [Sandia National Labs., Albuquerque, NM (United States); Blower, D.F.; Campbell, K.L. [Univ. of Michigan Transportation Research Institute, Ann Arbor, MI (United States). Center for National Truck Statistics

    1994-06-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. Several accident databases were reviewed to determine their suitability to the task. The TIFA (Trucks Involved in Fatal Accidents) database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports for selected TIFA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, fire temperature, fire size, fire separation, and fire duration.

  2. Proceedings of the workshop on severe accident research, Japan (SARJ-99)

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, Kazuichiro [ed.

    2000-11-01

    The Workshop on Severe Accident Research, Japan (SARJ-99) was taken place at Hotel Lungwood on November 8-10, 1999, and attended by 156 participants from 12 countries. A total of 46 papers, which covered wide areas of severe accident research both in experiments and analyses, such as fuel/coolant interaction, accident analysis and modeling, in-vessel phenomena, accident management, fission product behavior, research reactors, ex-vessel phenomena, and structural integrity, were presented. The panel discussion titled 'Link of Severe Accident Research Results to Regulation: Current Status and Future Perspective' was successfully conducted, and the wide variety of opinions and views were exchanged among panelists and experts. (J.P.N.)

  3. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  4. 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)

  5. Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

  6. Studies on melt-water-structure interaction during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Dinh, T.N.; Okkonen, T.J.; Bui, V.A.; Nourgaliev, R.R.; Andersson, J. [Royal Inst. of Technology, Div. of Nucl. Power Safety, Stockholm (Sweden)

    1996-10-01

    Results of a series of studies, on melt-water-structure interactions which occur during the progression of a core melt-down accident, are described. The emphasis is on the in-vessel interactions and the studies are both experimental and analytical. Since, the studies performed resulted in papers published in proceedings of the technical meetings, and in journals, copies of a set of selected papers are attached to provide details. A summary of the results obtained is provided for the reader who does not, or cannot, venture into the perusal of the attached papers. (au).

  7. Development of Highly Survivable Power and Communication System for NPP Instruments under Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seung J.; Gu, Beom W.; Nguyen, Duy T.; Choi, Bo H.; Rim, Chun T. [KAIST, Daejeon (Korea, Republic of); Lee, So I. [KHNP CRI, Daejeon (Korea, Republic of)

    2014-10-15

    According to the detail report from the Fukushima nuclear accident, the failure of conventional instruments is mainly due to the following reasons. 1) Insufficient backup battery capacity after the station black out (SBO) 2) The malfunction or damage of instruments due to the extremely harsh ambient condition after the severe accident 3) The cut-off of power and communication cable due to the physical shocks of hydrogen explosion after the severe accident Since the current equipment qualification (EQ) for the NPP instruments is based on the design basis accident such as loss of coolant accident (LOCA), conventional instruments, which are examined under EQ condition, cannot guarantee their normal operation during the severe accident. A 7m-long-distance wireless power transfer and a radio frequency (RF) communication were introduced with conventional wired system to increase a redundancy. A heat isolation box and a harness are adopted to provide a protection from the expected physical shocks such as missiles and drastic increase of ambient temperature and pressure. A detail design principle of the highly survivable power and communication system, which has 4 sub-systems of a DCRS wireless power transfer, a Zigbee wireless communication, a GFRP harness, and a passive type router with a fly back regulator, has been presented in this paper. Each sub-system has been designed to have a robust operation characteristic regardless of the estimated physical shocks after the severe accident.

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

    Science.gov (United States)

    Högberg, Lars

    2013-04-01

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

  9. An effect of containment filtered venting system on scale of contamination under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Ju young; Lee, Jai-ki [Hanyang Univ., Seoul (Korea, Republic of)

    2016-02-15

    Some countries are expected to expand the scope of the Emergency Planning Zone(EPZ) by the influence of Fukushima accident. However, if the equipment, which is able to mitigate the severe accident consequences, is installed, unnecessary costs for an expansion of emergency planning zone will be reduced. The International Nuclear Safety Advisory Group (INSAC) has suggested to mitigate severe accidents by installing The Filtered Containment Venting System (FCVS). The probabilistic assessment code MACCS2 was used to calculate the effective radiation dose with and without FCVS to determine the effective reduction by the installation of a FCVS.

  10. Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelm, Polina; Jobst, Matthias; Schaefer, Frank; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany)

    2016-05-15

    In the frame of the nuclear safety research program of the Helmholtz Association HZDR performs fundamental and applied research to assess and to reduce the risks related to the nuclear fuel cycle and the production of electricity in nuclear power plants. One of the research topics focuses on the safety aspects of current and future reactor designs. This includes the development and application of methods for analyses of transients and postulated accidents, covering the whole spectrum from normal operation till severe accident sequences including core degradation. This paper gives an overview of the severe accident research activities at the Reactor Safety Division at the Institute of Resource Ecology.

  11. Ruthenium behaviour in severe nuclear accident conditions. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Backman, U.; Lipponen, M.; Auvinen, A.; Jokiniemi, J.; Zilliacus, R. [VVT Processes (Finland)

    2004-08-01

    During routine nuclear reactor operations, ruthenium will accumulate in the fuel in relatively high concentrations. In a steam atmosphere, ruthenium is not volatile, and it is not likely to be released from the fuel. However, in an air ingress accident during reactor power operation or during maintenance, ruthenium may form volatile species, which may be released into the containment. Oxide forms of ruthenium are more volatile than the metallic form. Radiotoxicity of ruthenium is high both in the short and the long term. The results of this project imply that in oxidising conditions during nuclear reactor core degradation, ruthenium release increases as oxidised gaseous species Ru03 and Ru04 are formed. A significant part of the released ruthenium is then deposited on reactor coolant system piping. However, in the presence of steam and aerosol particles, a substantial amount of ruthenium may be released as gaseous Ru04 into the containment atmosphere. (au)

  12. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  13. The kinetics of aerosol particle formation and removal in NPP severe accidents

    Science.gov (United States)

    Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

    2016-06-01

    Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

  14. SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

    2001-07-01

    Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

  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. Incorporation of severe accidents in the licensing of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Alvarenga, Marco Antonio Bayout; Rabello, Sidney Luiz, E-mail: bayout@cnen.gov.b, E-mail: sidney@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN) Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Severe accidents are the result of multiple faults that occur in nuclear power plants as a consequence from the combination of latent failures and active faults, such as equipment, procedures and operator failures, which leads to partial or total melting of the reactor core. Regardless of active and latent failures related to the plant management and maintenance, aspects of the latent failures related to the plant design still remain. The lessons learned from the TMI accident in the U.S.A., Chernobyl in the former Soviet Union and, more recently, in Fukushima, Japan, suggest that severe accidents must necessarily be part of design-basis of nuclear power plants. This paper reviews the normative basis of the licensing of nuclear power plants concerning to severe accidents in countries having nuclear power plants under construction or in operation. It was addressed not only the new designs of nuclear power plants in the world, but also the design changes in plants that are in operation for decades. Included in this list are the Brazilian nuclear power plants, Angra-1, Angra-2, and Angra-3. This paper also reviews the current status of licensing in Brazil and Brazilian standards related to severe accidents. It also discusses the impact of severe accidents in the emergency plans of nuclear power plants. (author)

  17. CHEMICAL EFFECTS ON PWR SUMP STRAINER BLOCKAGE AFTER A LOSS-OF-COOLANT ACCIDENT: REVIEW ON U.S. RESEARCH EFFORTS

    Directory of Open Access Journals (Sweden)

    CHI BUM BAHN

    2013-06-01

    Full Text Available Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET, and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.

  18. Degraded core analysis for the PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  19. Prediction of structural integrity of steam generator tubes under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States)

    1999-11-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800 C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models. (orig.)

  20. Impact evaluation of the accident with release of a PWR coolant. Case study: Angra 3; Avaliacao do impacto de acidente com liberacao do refrigerante de reator PWR. Estudo de caso: Angra 3

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-10-26

    It was postulated in the cooling system, a LOCA where was lost 431 m{sup 3} of coolant. The inventory was 1.87 x 10{sup 10} Bq/m{sup 3} of tritium, 2.22 x 10{sup 7} Bp/m{sup 3} of cobalt and 3.48 x 10{sup 8} Bq/m{sup 3} of cesium and was launched near tue Itaorna beach, Angra dos Reis, RJ, Brazil. By applying the model in the proposed scenery (Angra 1 and 2 functioning and Angra 3 with variation of water taking and discharge with a progressive reduction after the accident), the dilution of specific activity of the radionuclides reached inferior values after 22 hours, to the reference values. After 54 hours, the levels of radionuclides, in the indirect influence are already below the minimum values of activity detected by the laboratory of environmental monitoring of the CNAAA

  1. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  2. Types and severity of operated supraclavicular brachial plexus injuries caused by traffic accidents.

    Science.gov (United States)

    Kaiser, Radek; Waldauf, Petr; Haninec, Pavel

    2012-07-01

    Brachial plexus injuries occur in up to 5% of polytrauma cases involving motorcycle accidents and in approximately 4% of severe winter sports injuries. One of the criteria for a successful operative therapy is the type of lesion. Upper plexus palsy has the best prognosis, whereas lower plexus palsy is surgically untreatable. The aim of this study was to evaluate a group of patients with brachial plexus injury caused by traffic accidents, categorize the injuries according to type of accident, and look for correlations between type of palsy (injury) and specific accidents. A total of 441 brachial plexus reconstruction patients from our department were evaluated retrospectively(1993 to 2011). Sex, age, neurological status, and the type and cause of injury were recorded for each case. Patients with BPI caused by a traffic accident were assessed in detail. Traffic accidents were the cause of brachial plexus injury in most cases (80.7%). The most common type of injury was avulsion of upper root(s) (45.7%) followed by rupture (28.2%), complete avulsion (16.9%) and avulsion of lower root(s) (9.2%). Of the patients, 73.9% had an upper,22.7% had a complete and only 3.4% had a lower brachial plexus palsy. The main cause was motorcycle accidents(63.2%) followed by car accidents (23.5%), bicycle accidents(10.7%) and pedestrian collisions (3.1%) (paccidents had a higher percentage of lower avulsion (22.7%) and a lower percentage of upper avulsion (29.3%), whereas cyclists had a higher percentage of upper avulsion (68.6%) based on the data from the entire group of patients (paccidents (9.3%,paccidents),significantly more upper and fewer lower palsies were present. In the bicycle accident group, upper palsy was the most common (89%). Study results indicate that the most common injury was an upper plexus palsy. It was characteristic of bicycle accidents, and significantly more common in car and motorcycle accidents. The results also indicate that it is important to consider the

  3. Insights into the behavior of nuclear power plant containments during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S.; Ludwigsen, J.S.; Parks, M.B.; Lambert, L.D. [Sandia National Labs., Albuquerque, NM (United States); Dameron, R.A.; Rashid, Y.R. [ANATECH Research Corp., San Diego, CA (United States)

    1993-06-01

    The containment building surrounding a nuclear reactor offers the last barrier to the release of radioactive materials from a severe accident into the environment. The loading environment of the containment under severe accident conditions may include much greater than design pressures and temperatures. Investigations into the performance of containments subject to ultimate or failure pressure and temperature conditions have been performed over the last several years through a program administered by the Nuclear Regulatory Commission (NRC). These NRC sponsored investigations are subsequently discussed. Reviewed are the results of large scale experiments on reinforced concrete, prestressed concrete, and steel containment models pressurized to failure. In conjunction with these major tests, the results of separate effect testing on many of the critical containment components; that is, aged and unaged seals, a personnel air lock and electrical penetration assemblies subjected to elevated temperature and pressure have been performed. An objective of the NRC program is to gain an understanding of the behavior of typical existing and planned containment designs subject to postulated severe accident conditions. This understanding has led to the development of experimentally verified analytical tools that can be applied to accurately predict their ultimate capacities useful in developing severe accident mitigation schemes. Finally, speculation on the response of containments subjected to severe accident conditions is presented.

  4. Integrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Yang, Ming; Zhang, Zhijian [Harbin Engineering University, Harbin (China)

    2014-08-15

    The development of a new risk monitor system is introduced in this paper, which can be applied not only to severe accident prevention in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The summary of the fundamental method is summarized on how to configure the Plant Defense in-Depth (Did) Risk Monitor by object-oriented software system based on functional modeling approach. Following the authors??preceding preliminary study for AP1000, the way of realizing the proposed method of configuring the plant Did risk monitor was investigated for a safety-enhanced Japanese PWR design to meet with the tight anti-severe accident requirements set by national regulation in Japan after Fukushima Daiichi accident. The result of this example practice of the presented preliminary study for Japanese PWR was for the level 4 of the Did in case of beyond design basis accident, that is, loss of all AC power + RCP seal LOCA, against the former case of AP1000 for level 3 Did in case of large LOCA.

  5. Risk factors associated with traffic violations and accident severity in China.

    Science.gov (United States)

    Zhang, Guangnan; Yau, Kelvin K W; Chen, Guanghan

    2013-10-01

    With the recent economic boom in China, vehicle volume and the number of traffic accident fatalities have become the highest in the world. Meanwhile, traffic accidents have become the leading cause of death in China. Systematically analyzing road safety data from different perspectives and applying empirical methods/implementing proper measures to reduce the fatality rate will be an urgent and challenging task for China in the coming years. In this study, we analyze the traffic accident data for the period 2006-2010 in Guangdong Province, China. These data, extracted from the Traffic Management Sector-Specific Incident Case Data Report, are the only officially available and reliable source of traffic accident data (with a sample size>7000 per year). In particular, we focus on two outcome measures: traffic violations and accident severity. Human, vehicle, road and environmental risk factors are considered. First, the results establish the role of traffic violations as one of the major risks threatening road safety. An immediate implication is: if the traffic violation rate could be reduced or controlled successfully, then the rate of serious injuries and fatalities would be reduced accordingly. Second, specific risk factors associated with traffic violations and accident severity are determined. Accordingly, to reduce traffic accident incidence and fatality rates, measures such as traffic regulations and legislation-targeting different vehicle types/driver groups with respect to the various human, vehicle and environment risk factors-are needed. Such measures could include road safety programs for targeted driver groups, focused enforcement of traffic regulations and road/transport facility improvements. Data analysis results arising from this study will shed lights on the development of similar (adjusted) measures to reduce traffic violations and/or accident fatalities and injuries, and to promote road safety in other regions. Copyright © 2013 Elsevier Ltd. All

  6. The Impact of Heat Waves on Occurrence and Severity of Construction Accidents

    Directory of Open Access Journals (Sweden)

    Rameez Rameezdeen

    2017-01-01

    Full Text Available The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers’ health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002–2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS policies.

  7. The Impact of Heat Waves on Occurrence and Severity of Construction Accidents

    Science.gov (United States)

    Rameezdeen, Rameez; Elmualim, Abbas

    2017-01-01

    The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers’ health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002–2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS) policies. PMID:28085067

  8. The Impact of Heat Waves on Occurrence and Severity of Construction Accidents.

    Science.gov (United States)

    Rameezdeen, Rameez; Elmualim, Abbas

    2017-01-11

    The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers' health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002-2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS) policies.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  10. OVERVIEW OF CONTAINMENT FILTERED VENT UNDER SEVERE ACCIDENT CONDITIONS AT WOLSONG NPP UNIT 1

    Directory of Open Access Journals (Sweden)

    Y.M. SONG

    2013-10-01

    Full Text Available Containment Filtered Vent Systems (CFVSs have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

  11. Evaluation of severe accident risks, Peach Bottom, Unit 2: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Payne, A.C.; Breeding, R.J.; Jow, H.N.; Shiver, A.W. (Sandia National Labs., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA)); Smith, L.N. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-12-01

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Peach Bottom Atomic Power Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Electric Company. The emphasis in this risk analysis was not on determining a so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events both internal and external to the power station were assessed. 39 refs., 174 figs., 133 tabs.

  12. 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

  13. Hot Cell Installation and Demonstration of the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [ORNL; Burns, Zachary M. [ORNL; Terrani, Kurt A. [ORNL; Yan, Yong [ORNL

    2017-08-01

    A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examine postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.

  14. Causes and Severity of Fatal Injuries in Autopsies of Victims of Fatal Traffic Accidents

    Directory of Open Access Journals (Sweden)

    F Panahi

    2010-03-01

    Full Text Available Introduction: In this retrospective study, we decided to determine the death causes and severity of injuries in traffic accidents according to reports of the forensic medical center of Yazd. Methods: A total of 251 fatalities due to traffic accidents that had undergone autopsy examinations at the Yazd forensic medicine center from2006 till 2008 were included in the study by census method. Data regarding gender, road user type, type of vehicle (car, motorcycle, autobus or minibus, consciousness level, and intensive care unit (ICU admission was gathered. For evaluation of injury severity, we used Injury Severity Score (ISS. Results: The population under study consisted of 202 men (80.5% and 49 women (19.5% with an average age of 34.1 years (range: 1-89 years. Motorcycle-pedestrian accidents were the most common type of injury (100, 39.8%. Head (220, 87.6% and face (169, 67.3% were the two most common sites of injuries. Mean (±SD of ISS was 23.2 (±10.4. According to autopsy records, the main cause of death was head trauma (146, 58.1%. Conclusion: Public awareness in terms of primary prevention of road accidents should be considered important. Also, regarding the high prevalence of brain injuries and complications associated with skull fractures, accessibility to neurosurgeons and availability of imaging devices have an important role in decreasing the mortality rate of traffic accidents.

  15. 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.

  16. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  17. 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.

  18. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu [Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki (Japan)

    2017-03-15

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

  19. Severe accident progression perspectives for Mark I containments based on the IPE results

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Drouin, M. [Nuclear Regulatory Commission, N. Bethesda, MD (United States)

    1995-12-31

    Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF.

  20. FN-curves: preliminary estimation of severe accident risks after Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Costa, Antonio Carlos Lopes da, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: aclc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    Doubts of whether the risks related to severe accidents in nuclear reactors are indeed very low were raised after the nuclear accident at Fukushima Daiichi in 2011. Risk estimations of severe accidents in nuclear power plants involve both probability and consequence assessment of such events. Among the ways to display risks, risk curves are tools that express the frequency of exceeding a certain magnitude of consequence. Societal risk is often represented graphically in a FN-curve, a type of risk curve, which displays the probability of having N or more fatalities per year, as a function of N, on a double logarithmic scale. The FN-curve, originally introduced for the assessment of the risks in the nuclear industry through the U.S.NRC Reactor Safety Study WASH-1400 (1975), is used in various countries to express and limit risks of hazardous activities. This first study estimated an expected rate of core damage equal to 5x10{sup -5} by reactor-year and suggested an upper bound of 3x10{sup -4} by reactor-year. A more recent report issued by Electric Power Research Institute - EPRI (2008) estimates a figure of the order of 2x10{sup -5} by reactor-year. The Fukushima nuclear accident apparently implies that the observed core damage frequency is higher than that predicted by these probabilistic safety assessments. Therefore, this paper presents a preliminary analyses of the FN-curves related to severe nuclear reactor accidents, taking into account a combination of available data of past accidents, probability modelling to estimate frequencies, and expert judgments. (author)

  1. Sensitivity study for accident tolerant fuels: Property comparisons and behavior simulations in a simplified PWR to enable ATF development and design

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Kristina Yancey, E-mail: kristina.yancey@gmail.com; Sudderth, Laura; Brito, Ryan A.; Evans, Jordan A.; Hart, Clifford S.; Hu, Anbang; Jati, Andi; Stern, Karyn; McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu

    2016-12-01

    Highlights: • This study compared four accident tolerant fuels against uranium dioxide. • Material property correlations were developed to evaluate fuel performance. • The fuels’ neutronic and thermal hydraulic behaviors were studied in the AP1000. • No fuel type performed better in all areas, but each has strengths and weaknesses. • More research is needed to build a complete model of the fuel performances. - Abstract: Since the events at the Fukushima-Daiichi nuclear power plant, there has been increased interest in developing fuels to better withstand accidents for current light water reactors. Four accident tolerant fuel candidates are uranium oxide with beryllium oxide additives, uranium oxide with silicon carbide matrix additives, uranium nitride, and uranium nitride with uranium silicide composite. The first two candidates represent near-term high performance uranium oxide with high thermal conductivity and neutron transparency, and the second two represent mid-term high-density fuels with highly beneficial thermal properties. This study seeks to understand the benefits and drawbacks of each option in place of uranium dioxide. To assess the material properties for each of the fuel types, an extensive literature review was performed for material property data. Correlations were then made to evaluate the properties during reactor operation. Neutronics and thermal hydraulics studies were also completed to determine the impact of the use of each candidate in an AP1000 reactor. In most cases, the candidate fuels performed more desirably than uranium dioxide, but no fuel type performed better in all aspects. Much more research needs to be performed to build a complete model of the fuel performances, primarily experimental data for uranium silicide. Each of the fuels studied has its own benefits and drawbacks, and the comparisons discussed in this report can be used to aid in determining the most appropriate fuel depending on the desired specifications.

  2. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

  5. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  6. Risk factors associated with bus accident severity in the United States: A generalized ordered logit model

    DEFF Research Database (Denmark)

    Kaplan, Sigal; Prato, Carlo Giacomo

    2012-01-01

    Introduction: Recent years have witnessed a growing interest in improving bus safety operations worldwide. While in the United States buses are considered relatively safe, the number of bus accidents is far from being negligible, triggering the introduction of the Motor-coach Enhanced Safety Act...... that accident severity increases: (i) for young bus drivers under the age of 25; (ii) for drivers beyond the age of 55, and most prominently for drivers over 65 years old; (iii) for female drivers; (iv) for very high (over 65 mph) and very low (under 20 mph) speed limits; (v) at intersections; (vi) because...

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

    Energy Technology Data Exchange (ETDEWEB)

    Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

    1995-09-01

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

  8. Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, Joy Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knudson, Darrell Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented

  9. Severe Accidents and New Reactors. Twenty Years of Research; Accidents severos y nuevos reactores. Veinte anos de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.

    2008-07-01

    A review was done on the main activities performed by the Programme for Nuclear Safety of CIEMAT in the field of nuclear reactor safety from 1985 to 2005. It covers the areas of severe accident and source term, advanced and passive reactors, containments analyses and plant applications. It is emphasized CIEMATs participation in national and international projects mainly in those supported by CSN, OECD and the EU. At the same time, experimental and analytical capabilities set up at CIEMAT, as PECA, RECA and GIRS for simulating aerosol pool scrubbing phenomena, hydrogen catalytic recombiner and sprays are been presented, together with an Annex on Generation IV. Two chapters were added, one on the nuclear power reactors in the world and another about the safety systems and principles. (Author)

  10. Incorporating real-time traffic and weather data to explore road accident likelihood and severity in urban arterials.

    Science.gov (United States)

    Theofilatos, Athanasios

    2017-06-01

    The effective treatment of road accidents and thus the enhancement of road safety is a major concern to societies due to the losses in human lives and the economic and social costs. The investigation of road accident likelihood and severity by utilizing real-time traffic and weather data has recently received significant attention by researchers. However, collected data mainly stem from freeways and expressways. Consequently, the aim of the present paper is to add to the current knowledge by investigating accident likelihood and severity by exploiting real-time traffic and weather data collected from urban arterials in Athens, Greece. Random Forests (RF) are firstly applied for preliminary analysis purposes. More specifically, it is aimed to rank candidate variables according to their relevant importance and provide a first insight on the potential significant variables. Then, Bayesian logistic regression as well finite mixture and mixed effects logit models are applied to further explore factors associated with accident likelihood and severity respectively. Regarding accident likelihood, the Bayesian logistic regression showed that variations in traffic significantly influence accident occurrence. On the other hand, accident severity analysis revealed a generally mixed influence of traffic variations on accident severity, although international literature states that traffic variations increase severity. Lastly, weather parameters did not find to have a direct influence on accident likelihood or severity. The study added to the current knowledge by incorporating real-time traffic and weather data from urban arterials to investigate accident occurrence and accident severity mechanisms. The identification of risk factors can lead to the development of effective traffic management strategies to reduce accident occurrence and severity of injuries in urban arterials. Copyright © 2017 Elsevier Ltd and National Safety Council. All rights reserved.

  11. Risk factors affecting the severity of traffic accidents at Shanghai river-crossing tunnel.

    Science.gov (United States)

    Lu, Jian John; Xing, Yingying; Wang, Chen; Cai, Xiaonan

    2016-01-01

    With increasing traffic volume and urban development, increasing numbers of underground tunnels have been constructed to relieve conflict between strained land and heavy traffic. However, as more long tunnels are constructed, tunnel traffic safety is becoming increasingly serious. Thus, it is necessary to acquire their implications and impacts. This study examined 4,539 traffic accidents that have occurred in 14 Shanghai river-crossing tunnels for the period 2011-2012 and analyze the correlation between potential factors and accident injury severity. An ordered logit model was developed to examine the correlation between potential factors and accident injury severity. Results show that increased injury severity is associated with male drivers, drivers aged 65 years or older, accident time from midnight to dawn, weekends, wet road surface, goods vehicles, 3 or more vehicles, 4 or more lanes, middle speed limits (50-79 km/h), zone 3, extra-long tunnels (over 3,000 m), and maximum longitudinal gradient. This article aims to provide useful information for engineers to develop interventions and countermeasures to improve tunnel safety in China.

  12. Insoluble aerosol behavior inside the PCCS condenser tube under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, A.; Nemoto, K.; Akinaga, M. [Toshiba Corp., Kawasaki (Japan); Oikawa, H. [Toshiba Corp., Yokohama (Japan)

    1996-07-01

    The passive containment cooling system (PCCS), which has been incorporated into the advanced light water reactor (ALWR) design, has the capability of post accident decay heat removal by means of natural force driven condensation heat transfer. Since some uncertainties remain in the PCCS performance during a severe accident especially in the amount of aerosol deposition which causes the heat transfer degradation, the experiment had been performed previously simulating single condenser tube, postulated steam and noncondensable gas flow rate using prototypical soluble aerosol (CsI). The observed aerosol deposition rate onto the condenser tube surface was quite small under steam rich condition. However, during the severe accident, insoluble aerosols such as structural material might also be released and flow into the PCCS as well as soluble aerosol, and the deposition behavior has not been clarified. Thus, the experiment using a polystyrene LATEX was conducted under the same conditions in which the soluble aerosol test was performed. The experimental results showed similar trend as that of the soluble aerosol case, and especially in case of steam rich condition, the amount of deposition was below detection limit. The deposition rate in other cases are consistent with the prediction by existing theoretical correlation. Analytical sensitivity study varying inlet flow condition indicated no significant increase of aerosol deposition. These results suggest promising performance of PCCS under severe accident condition.

  13. Supported Pd nanoclusters for the hydrogen mitigation application in severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Zhengfeng; Zhu, Hongzhi; Zhang, Zhi; Zheng, Zhenghua [China Academy of Engineering Physics, P. O. Box 919-71, Mianyang 621700 (China); Ma, Guohua [University of Science and Technology of Southwest, Mianyang 621010 (China); Lai, Xinchun; Li, Rong; Tang, Tao; Fu, Jun [China Academy of Engineering Physics, P. O. Box 919-71, Mianyang 621700 (China); Gao, Bo, E-mail: gaobo@caep.cn [China Academy of Engineering Physics, P. O. Box 919-71, Mianyang 621700 (China)

    2017-05-15

    Highlights: • Pd catalysts were prepared by electroless deposition path with no extra reduction agents. • The Pd catalysts not only have good hydrogen-oxygen recombination efficiency, but also have good stability. • The catalysts were proved to have good resistance to poisoning. • Pd catalysts could be supposed to be used for PARs in severe accidents. - Abstract: Accidents at TMI, USA and Fukushima, Japan have emphasized the need for hydrogen mitigation during nuclear plant accidental conditions, especially during severe accidents which will be no power, massive hydrogen, high temperature, long-term operation, and poisoning environment. Passive autocatalytic recombiners with catalyst sheets are the promising way to deal with the situation in severe accidents. Here we report a new kind of catalyst sheets based on stainless steel supported Pd nanoclusters prepared by electroless deposition route. The catalyst sheets were characterised for morphology and composition of surface by SEM and EDS. The catalytic activity of the catalyst sheets has been evaluated under the conditions of higher temperature, long-term operation and poisoning environments. The catalyst sheets showed high activity and good stability either operating above 500 °C for 24 h or continuous operating for 25 days. For the obtained catalyst sheets after exposed to methanal, iodine vapor and BaSO{sub 4} aerosol respectively with corresponding concentrations higher than SA conditions, the start-up time for H{sub 2}-O{sub 2} recombination reaction was less than 1 min and the catalytic efficiency was more than 90%. These results indicate the potential application of this type of catalyst sheets for hydrogen mitigation in severe accidents.

  14. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  15. A Basic Study on the Ejection of ICI Nozzle under Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jong Rae; Bae, Ji Hoon; Bang, Kwang Hyun [Korea Maritime and Ocean University, Busan (Korea, Republic of); Park, Jong Woong [Dongguk University, Gyeongju (Korea, Republic of)

    2016-05-15

    Nozzle injection should be blocked because it affect to the environment if its melting core exposes outside. The purpose of this study is to carry out the thermos mechanical analysis due to debris relocation under severe accidents and to predict the nozzle ejection calculated considering the contact between the nozzle and lower head, and the supports of pipe cables. As a result of analyzing process of severe accidents, there was melting reaction between nozzle and the lower head. In this situation, we might predict the non-uniform contact region of nozzle hole of lower head and nozzle outside, delaying ejection of nozzles. But after melting, the average remaining length of the nozzle was 120mm and the maximum vertical displacement of lower nozzle near the weld is 3.3mm so there would be no nozzle this model, because the cable supports restrains the vertical displacement of nozzle.

  16. Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

    Science.gov (United States)

    Yamaji, Akifumi; Li, Xin

    2016-08-01

    Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

  17. ESTIMATION OF WEIBULL PARAMETERS USING A RANDOMIZED NEIGHBORHOOD SEARCH FOR THE SEVERITY OF FIRE ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Soontorn Boonta

    2013-01-01

    Full Text Available In this study, we applied Randomized Neighborhood Search (RNS to estimate the Weibull parameters to determine the severity of fire accidents; the data were provided by the Thai Reinsurance Public Co., Ltd. We compared this technique with other frequently-used techniques: the Maximum Likelihood Estimator (MLE, the Method of Moments (MOM, the Least Squares Method (LSM and the weighted least squares method (WLSM and found that RNS estimates the parameters more accurately than do MLE, MOM, LSM or WLSM."

  18. Prospect about hydrogen release at severe accident using mass gain from cladding oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Joon Young; Lee, Jae Young; Park, Sang Gil [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In order to understand the behavior of the fuel cladding during a severe accident, it is necessary to investigate the behavior of Zircaloy-4 in various conditions. Furthermore the oxidation by steam-nitrogen gas mixtures needs to be focused since severe accident at spent fuel pools such as partial LOCA scenario could make steam contained atmosphere and then causes different results compared to the oxidation by pure air. Therefore the experiments were conducted for depicting the Partial LOCA. The experimental condition is established by the acceptable range of experimental device, thermo-gravimetry (TG). This paper is based on a revised and considerably extended presentation given at the 21{sup st} International Quench Workshop. The hydrogen absorption is observed in the case of SF1{sub 1}100C supplied by the lowest flow rate (1 lpm) of the reactive gas composed by steam-nitrogen mixture. To confirm the role of the hydrogen absorption in the Zircaloy, precise measurement methods for metallography are needed. Through these experiments, it could be contributed to trace the amount of the hydrogen release from the cladding oxidation at severe accident.

  19. Development of a MAAP-based Severe Accident Training Simulator using Visual System Analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jae Seung [SENTECH, Daejeon (Korea, Republic of); Park, Soo Yong; Ahn, Kwang Il; Kim, Kyung Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The recent environment of severe accident analysis requires high performance computers to simulate complicated reactor and containment phenomena. In parallel with this, rapid advances in computer technology now enable these codes to run in real or almost real time. The remaining limitation restricting their use on an even wider scale is that most of the existing codes are still subject to a complicated I/O structure. Even user-friendly graphical user interfaces (GUI) will be not likely to help better and efficient interpretation of the analysis results obtained from these codes as well as for their increased use. This situation has motivated the development of easy-to-use GUI tools for severe accident codes, such as ViSA, SNAP, MAAP4-GRAAP, SATS, and et al. For instance, ViSA enables the thermal-hydraulic system codes to be used like a conventional nuclear plant analyzer. Recently, a project for a real time simulation of results obtained using MAAP4 codes under the ViSA environment has been initiated in KAERI. Such a GUI-based interactive interface can be very useful in sharing real time analysis results obtained from the MAAP code. The purpose of this paper is to introduce the current status of a MAAP-based Severe Accident Simulator being coupled with the ViSA system

  20. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H

    2007-04-15

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability.

  1. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K. J.; Park, C. K.; Seok, S. D.; Park, R. J.; Yi, S. J.; Kang, K. H.; Ham, Y. S.; Cho, Y. R.; Kim, J. H.; Jeong, J. H.; Shin, K. Y.; Cho, J. S.; Kim, D. H.

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  2. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  3. Recent severe accident research synthesis of the major outcomes from the SARNET network

    Energy Technology Data Exchange (ETDEWEB)

    Van Dorsselaere, J.-P., E-mail: jean-pierre.van-dorsselaere@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Auvinen, A. [VTT Technical Research Centre, Espoo (Finland); Beraha, D. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Chatelard, P. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Herranz, L.E. [Centro de Investigaciones Energéticas MedioAmbientales y Tecnológicas (CIEMAT), Madrid (Spain); Journeau, C. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Paris (France); Klein-Hessling, W. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Kljenak, I. [Jozef Stefan Institute (JSI), Ljubljana (Slovenia); Miassoedov, A. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Paci, S. [University of Pisa, Pisa (Italy); Zeyen, R. [European Commission Joint Research Centre, Institute for Energy (JRC/IET), Petten (Netherlands)

    2015-09-15

    Highlights: • SARNET network of excellence integration mid-2013 in the NUGENIA Association. • Progress of knowledge on corium behaviour, hydrogen explosion and source term. • Further development of ASTEC integral code to capitalize knowledge. • Ranking of next R&D high priority issues accounting for international research. • Dissemination of knowledge through education courses and ERMSAR conferences. - Abstract: The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core–concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented.

  4. Severe immune dysfunction after lethal neutron irradiation in a JCO nuclear facility accident victim.

    Science.gov (United States)

    Nagayama, Hitomi; Ooi, Jun; Tomonari, Akira; Iseki, Tohru; Tojo, Arinobu; Tani, Kenzaburo; Takahashi, Tsuneo A; Yamashita, Naohide; Shigetaka, Asano

    2002-08-01

    The optimal treatment for the hematological toxicity of acute radiation syndrome (ARS) is not fully established, especially in cases of high-dose nonuniform irradiation by mixed neutrons and gamma-rays, because estimation of the irradiation dose (dosimetry) and prediction of autologous hematological recovery are complicated. For the treatment of ARS, we performed HLA-DRB1-mismatched unrelated umbilical cord blood transplantation (CBT) for a nuclear accident victim who received 8 to 10 GyEq mixed neutron and gamma-ray irradiation at the JCO Co. Ltd. nuclear processing facility in Tokaimura, Japan. Donor/ recipient mixed chimerism was attained; thereafter rapid autologous hematopoietic recovery was achieved in concordance with the termination of immunosuppressants. Immune function examined in vitro showed recovery of the autologous immune system was severely impaired. Although the naive T-cell fraction and the helper T-cell subtype 1 fraction were increased, the mitogenic responses of T-cells and the allogeneic mixed leukocyte reaction were severely suppressed. Endogenous immunoglobulin production was also suppressed until 120 days after the accident. Although skin transplantation for ARS was successful, the patient died of infectious complications and subsequent acute respiratory distress syndrome 210 days after the accident. These results suggest that fast neutrons in doses higher than 8 to 10 Gy cause complete abrogation of the human immune system, which may lead to fatal outcome even if autologous hematopoiesis recovers. The roles of transplantation, autologous hematopoietic recovery, chimerism, immune suppression, and immune function are discussed.

  5. Evaluating the Effectiveness of Alternate Entry Condition into the Severe Accident Management Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyung Seok; Lee, Su Won [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Min, Shin Jung [Korea Hydro and Nuclear Power Co. Ltd. Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the effectiveness of the CA as an alternate means is evaluated quantitatively by utilizing the Modular Accident Analysis Program (MAAP) 5 computer code including the MAAP5-DOSE module, which can analyze the radiation level inside the containment. The effectiveness of the CA has been investigated by utilizing the MAAP5 code including the MAAP5- DOSE. The onset of core damage is considered to be a core (fuel rod cladding) condition at the time when the core exit temperature reaches the value prescribed for transition to Severe Accident Management Guidance (SAMG), which is 1200 .deg. F. However, during a shutdown state, the core exit thermocouples measurements are unavailable after lifting reactor vessel head. Thus, an alternate means to detect the onset of core damage is necessary to cover all plant operating states. In order for that, a Computational Aid (CA), 'Radiation Level as a Functional of Time after Shutdown,' has been developed. The upper containment radiation instrumentation is a gross gamma monitor, and has a reliable instrumentation range during severe accidents. It can be used for detecting onset of core damage. Thus, the radiation level can be used as alternative means of the entry condition into the SAMG. It has been shown that the SAMG entry timings determined by using the core exit thermocouple measurements and by the radiation monitoring with the CA would not be differentiated. The time difference estimates entering SAMG would be less 15 min which would not influence the operator action significantly.

  6. Severe Tricuspid Regurgitation Diagnosed 13 Years after a Car Accident: A Case Report

    Directory of Open Access Journals (Sweden)

    Burak Acar

    2015-10-01

    Full Text Available Blunt chest traumas mostly occur due to car accidents and can cause many cardiac complications such as septal rupture, free-wall rupture, coronary artery dissection or thrombosis, heart failure, arrhythmias, and chordae and papillary muscle rupture. One of the most serious complication is tricuspid regurgitation (TR, which can be simply diagnosed by physical examination and confirmed by echocardiography. We describe a 48-year-old female patient, diagnosed with severe TR 13 years after a blunt chest trauma due to a car accident. TR was diagnosed with transthoracic echocardiography and three dimensional transthoracic echocardiography had defined the exact pathology of the tricuspid valve. The patient underwent successful surgery with bioprosthetic valve implantation and was discharged at 6th postoperative day without any complication. The patient had no problem according to the follow-up one month and six months after operation

  7. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  8. Integrated hydrogen control solutions for severe accidents using passive autocatalytic recombiners

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, M.; Tietsch, W.; Sabate Farnos, R.

    2012-07-01

    In a severe accident or a beyond-design-basis-accident, the reaction of water with zirconium alloy cladding, radiolysis of water, corium-concrete reactions and other corrosion phenomena generate hydrogen (H2). The detonation of this H2 in containment or in auxiliary buildings can result in damage to structures or loss of containment integrity. Identifying the generation and special distribution of hydrogen and controlling its concentration with Passive Autocatalytic Recombiners (PARs) solves this concern. Westinghouse's approach for hydrogen management starts by defining the quantities and transport/distribution of H{sub 2} in-containment and out of containment with analysis tools such as MAAP, MELCOR, GASFLOW or FATE. Based on the results of these analyses, an optimized H2 Control Strategy is proposed in terms of number and location of PARs, and efficient integration with other H{sub 2} management devices like e.g. existing igniters, H{sub 2} monitors, etc.

  9. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  10. A retrspective study of rescuing severe open craniocerebral injuries caused by traffic accidents

    Institute of Scientific and Technical Information of China (English)

    陈长才; 宁可; 等

    1999-01-01

    Objective:To investigate the rescuing principles of severe open craniocerebral injuries caused by traffic accidents.Methods:A retrospective study was performed for 36 patients admitted to our hospital from January 1986 to December 1995,who suffered from severe open craniocerebral injuries in traffic accidents.Results:These 36 cases occupied 52.10% of all the severe open craniocerebral injuries during the same period.The clinical features included confusion of consciousness, extensive cerebral contusion and laceration,severe contamination of the wound,high incidence of intracranial hematoma and multiple system injuries.Nineteen patients.(63.34%)ecovered normal neurological function,7 were (23.33%)mild disabled,4(13.33%)severe disabled,2(5.56%) vegetative survival,and 4(11.11%)dead.Conclusions:The main principles of salvage should emphasize the importance of emergent prehospital rescue,and be transfered to a specialized hospital as soon as possible.Postoperative complications included severe brain edema,intracerebral infection,and pneumonia,Debriding thoroughly at early stage and treating complications effectively would lower the rate of mortality and disability.

  11. 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.

  12. Development of heat insulation device to protect pressure measuring instruments from high temperature under the severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Jaehyun; Shin, Sung Min; Kang, Hyun Gook [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Micro Control Unit (MCU), communication module, and power supply system are also needed to be protected for the pressure transmitter. The harsh condition in containment which is created by the severe accident are composed of five elements: high temperature, high pressure, high humidity, high radiation, and physical threats by shrapnel generated during the process of the severe accident. Among these five elements, high temperature should be focused because other elements can be solved even with the thin shield. In this study, a detailed design of the heat insulation device which will be installed in the containment based on the Min Yoo's study and a verification test are done. Development of heat insulation device which enables operator to get in-containment data for the proper mitigation process under the severe accident was done in this study. With researches for severe accident management systems which proceeding actively since the Fukushima accident, researches for reliable instrumentations of in-containment data which is necessary to operate severe accident management systems properly in harsh condition during accident also should be progressed.

  13. R and D relative to the serious accidents in the PWR type reactors: assessment and perspectives; R and D relative aux accidents graves dans les reacteurs a eau pressurisee: bilan et perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Bentaib, A.; Bonneville, H.; Caroli, H.; Chaumont, B.; Clement, B.; Cranga, M.; Koundy, V.; Laurent, B.; Micaelli, J.C.; Meignen, R.; Pichereau, F.; Plassart, D.; Van-Dorsselaere, P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Ducros, G.; Journeau, Ch.; Magallon, D. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Durin, M.; Studer, E. [CEA Saclay 91 - Gif sur Yvette (France); Seiler, J.M. [CEA Grenoble, 38 (France); Ranval, W. [Electricite de France (EDF), 75 - Paris (France)

    2006-07-01

    This document presents the current state of the research relative to the grave accidents realized in France and abroad. It aims at giving the most exhaustive possible and objective vision of this original field of research. He allows to contribute to the identification and to the hierarchical organization of the needs of R and D, this hierarchical organization in front of, naturally, to be completed by a strong lighting on needs in terms of safety analyses associated with the different risks and the physical phenomena, in particular with the support of probability evaluations of safety level 2, whose the level of sharpness must be sufficient not to hide, by construction, physical phenomena of which the limited knowledge leads to important uncertainties. Let us note that neither the safety analyses, nor the E.P.S. 2 are presented in this document. This report presents the physical phenomena which can arise during a grave accident, in the reactor vessel and in the reactor containment, their chain and the means allowing to ease the effects. The corresponding scenarios are presented to the chapter 2. The chapter 3 is dedicated to the progress of the accident in the reactor vessel; the degradation of the core in reactor vessel (3.1), the behavior of the corium in bottom of reactor vessel (3.2) the break of the reactor vessel (3.3) and the fusion in pressure (3.4) are thus handled there. The chapter 4 concerns the phenomena which can lead to a premature failure of the containment, namely the direct heating of gases of the containment (4.1), the hydrogen risk (4.2) and the vapor explosion (4.3). The phenomenon which can lead to a delayed failure from the containment, namely the interaction corium-concrete, is approached on the chapter 5. The chapter 6 is dedicated to the problems connected to the keeping back and to the corium cooling in reactor vessel and out of reactor vessel, namely the keeping back in reactor vessel by re-flooding of the primary circuit or by re

  14. 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.

  15. Prediction of hydrogen concentration in nuclear power plant containment under severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Kim, Dong Yeong; Yoo, Kwae Hwan; Na, Man Gyun, E-mail: magyna@chosun.ac.kr

    2016-04-15

    Highlights: • We present a hydrogen-concentration prediction method in an NPP containment. • The cascaded fuzzy neural network (CFNN) is used in this prediction model. • The CFNN model is much better than the existing FNN model. • This prediction can help prevent severe accidents in NPP due to hydrogen explosion. - Abstract: Recently, severe accidents in nuclear power plants (NPPs) have attracted worldwide interest since the Fukushima accident. If the hydrogen concentration in an NPP containment is increased above 4% in atmospheric pressure, hydrogen combustion will likely occur. Therefore, the hydrogen concentration must be kept below 4%. This study presents the prediction of hydrogen concentration using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model was developed using data on severe accidents in NPPs. The data were obtained by numerically simulating the accident scenarios using the MAAP4 code for optimized power reactor 1000 (OPR1000) because real severe accident data cannot be obtained from actual NPP accidents. The root-mean-square error level predicted by the CFNN model is below approximately 5%. It was confirmed that the CFNN model could accurately predict the hydrogen concentration in the containment. If NPP operators can predict the hydrogen concentration in the containment using the CFNN model, this prediction can assist them in preventing a hydrogen explosion.

  16. Effective Factors in Severity of Traffic Accident-Related Traumas; an Epidemiologic Study Based on the Haddon Matrix.

    Science.gov (United States)

    Masoumi, Kambiz; Forouzan, Arash; Barzegari, Hassan; Asgari Darian, Ali; Rahim, Fakher; Zohrevandi, Behzad; Nabi, Somayeh

    2016-01-01

    Traffic accidents are the 8(th) cause of mortality in different countries and are expected to rise to the 3(rd) rank by 2020. Based on the Haddon matrix numerous factors such as environment, host, and agent can affect the severity of traffic-related traumas. Therefore, the present study aimed to evaluate the effective factors in severity of these traumas based on Haddon matrix. In the present 1-month cross-sectional study, all the patients injured in traffic accidents, who were referred to the ED of Imam Khomeini and Golestan Hospitals, Ahvaz, Iran, during March 2013 were evaluated. Based on the Haddon matrix, effective factors in accident occurrence were defined in 3 groups of host, agent, and environment. Demographic data of the patients and data regarding Haddon risk factors were extracted and analyzed using SPSS version 20. 700 injured people with the mean age of 29.66 ± 12.64 years (3-82) were evaluated (92.4% male). Trauma mechanism was car-pedestrian in 308 (44%) of the cases and car-motorcycle in 175 (25%). 610 (87.1%) cases were traffic accidents and 371 (53%) occurred in the time between 2 pm and 8 pm. Violation of speed limit was the most common violation with 570 (81.4%) cases, followed by violation of right-of-way in 57 (8.1%) patients. 59.9% of the severe and critical injuries had occurred on road accidents, while 61.3% of the injuries caused by traffic accidents were mild to moderate (p accidents (p traffic accident-related traumas were age over 50, not using safety tools, and undertaking among host-related factors; insufficient environment safety, road accidents and time between 2 pm and 8 pm among environmental factors; and finally, rollover, car-pedestrian, and motorcycle-pedestrian accidents among the agent factors.

  17. Use of detailed thermochemical databases to model chemical interactions in the Severe Accident codes

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M. [IPSN/DRS, CEA Cadarache (France)

    2001-07-01

    For the prevention, mitigation and management of severe accidents, many problems related to core melt have to be solved: fuel degradation, melting and relocation, convection in the core melt(s), coolability of the core melt(s), fission product release, hydrogen production, behavior of the materials of the protective layers, ex-vessel spreading of the core melt(s).. To solve these problems such properties like thermal conductivity, heat capacity, density, viscosity, evaporation or sublimation of melts, the solidification behavior (solid/liquid fraction), the tendency to trap or to release the fission products, the stratification of melts notably metallic and oxide, must be known. However most of these properties are delicate to measure directly at high temperature and/or in the radio-active environment produced by the fission products. Therefore some of them must be derived by calculations from the physical-chemical description of the melt: number of phases, phase compositions, proportions of solids and liquids and their respective oxidation state, miscibility of the liquids, solubility of one phase in another, etc. This information is given by the phase diagrams of the materials in presence. Since more than ten years, IPSN has developed in collaboration with THERMODATA (Grenoble, France) a very detailed thermochemical database for the complex system U-O-Zr-Fe-Ni-La-Ba-Ru-Sr-Si-Mg-Ca-Al-(H-Ar). The direct coupling between the severe accident (SA) Codes and a thermochemical code with its database is not actually possible because of the computer time consuming and the size of the database. For this reason, most of the Severe Accident codes usually have a very simplified description for the phase diagrams which are not in agreement with the status of the art. In this presentation, alternative methodologies are detailed with their respective difficulties, the goal being to build an interface between a thermochemical database and a SA Code and to get a fast, accurate and

  18. Severe accident improvements for Carem-25 to arrest reactor vessel meltdown sequences

    Energy Technology Data Exchange (ETDEWEB)

    Poier Baez, L.E.; Nunez Mac Leod, J.E.; Baron, J.H. [Cuyo National University, Engineering Faculty, Mendoza (Argentina)

    2001-07-01

    Given an accident sequence, that leads to sustained uncovering of the core, the progression of core damage involves several complex phenomena. The progression of these phenomena can lead to a breach of the reactor vessel followed by the discharge of molten core materials to the containment. Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the first analyses, via numerical simulation, for the conceptual design of such a container type; furthermore, the paper addresses simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. MELCOR is a fully integrated relatively fast running code that models the progression of accidents in light water reactor power plants. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher. (authors)

  19. Severe accident improvements for Carem-25 to arrest reactor vessel meltdown sequences

    Energy Technology Data Exchange (ETDEWEB)

    Poier Baez, L.E.; Nunez Mac Leod, J.E.; Baron, J.H. [Cuyo National University, Engineering Faculty, Mendoza (Argentina)

    2001-07-01

    Given an accident sequence, that leads to sustained uncovering of the core, the progression of core damage involves several complex phenomena. The progression of these phenomena can lead to a breach of the reactor vessel followed by the discharge of molten core materials to the containment. Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the first analyses, via numerical simulation, for the conceptual design of such a container type; furthermore, the paper addresses simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. MELCOR is a fully integrated relatively fast running code that models the progression of accidents in light water reactor power plants. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher. (authors)

  20. Workshop proceedings of ISAMM 2009: Implementation of severe accident management measures

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, S. (ed.) [Paul Scherrer Institute (PSI), Nuclear Energy and Safety Research Department, Laboratory for Thermal Hydraulics, ViIligen (Switzerland)

    2010-10-15

    This comprehensive report published by the Paul Scherrer Institute (PSI) in Switzerland reports on a conference and workshop held in Switzerland in October 2009 dealing with Severe Accidents Management (SAM) in nuclear power stations. The workshop provided an update on the status of severe accident management measures and their implications since the OECD/CSNI workshop held in 2001 at the PSI in Switzerland. Since the 2001 workshop, additional work has been performed to integrate emergency procedures and SAM measures into risk assessments in order to better reflect operator responses to recover a plant from a damaged state. The major focus of the workshop was to address SAM measures for both operational plants and new plant designs. Also, the integration of SAM measures into contemporary/future probabilistic risk assessments was discussed. 41 papers were presented in 8 sessions. The papers addressed the following areas: 1) Current status and insights of SAM (2 sessions); 2) Probabilistic Safety Assessment (PSA) modelling issues; 3) code analysis for supporting Serious Accident Management Guidance (SAMG, 2 sessions); 4) decision making, tools, training, risk-targets and entrance to SAM; 5) design modifications for implementation of SAM; 6) physical phenomena. The last part of the workshop was devoted to the presentation of the most striking highlights of the papers in the above areas, followed by two panellists giving presentations on human and organisational aspects of SAM, their importance in relation to technical issues and the effectiveness of current SAMG implementation. The question of how consequence analyses can be used to improve the effectiveness of SAM is discussed. The contributions were presented by representatives from Austria, Germany, Japan, France, the USA, Korea, Switzerland, Finland, Hungary, Belgium, Canada, Sweden, the Czech republic, the United kingdom, the Netherlands, Spain, Slovenia and Russia. The authors state that the overall picture

  1. Safety Implementation of Hydrogen Igniters and Recombiners for Nuclear Power Plant Severe Accident Management

    Institute of Scientific and Technical Information of China (English)

    XIAO Jianjun; ZHOU Zhiwei; JING Xingqing

    2006-01-01

    Hydrogen combustion in a nuclear power plant containment building may threaten the integrity of the containment. Hydrogen recombiners and igniters are two methods to reduce hydrogen levels in containment buildings during severe accidents. The purpose of this paper is to evaluate the safety implementation of hydrogen igniters and recombiners. This paper analyzes the risk of deliberate hydrogen ignition and investigates three mitigation measures using igniters only, hydrogen recombiners only or a combination of recombiners and igniters. The results indicate that steam can effectively control the hydrogen flame acceleration and the deflagration-to-detonation transition.

  2. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    Science.gov (United States)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  3. Longitudinal Associations Between PTSD Symptoms and Dyadic Conflict Communication Following a Severe Motor Vehicle Accident.

    Science.gov (United States)

    Fredman, Steffany J; Beck, J Gayle; Shnaider, Philippe; Le, Yunying; Pukay-Martin, Nicole D; Pentel, Kimberly Z; Monson, Candice M; Simon, Naomi M; Marques, Luana

    2017-03-01

    There are well-documented associations between posttraumatic stress disorder (PTSD) symptoms and intimate relationship impairments, including dysfunctional communication at times of relationship conflict. To date, the extant research on the associations between PTSD symptom severity and conflict communication has been cross-sectional and focused on military and veteran couples. No published work has evaluated the extent to which PTSD symptom severity and communication at times of relationship conflict influence each other over time or in civilian samples. The current study examined the prospective bidirectional associations between PTSD symptom severity and dyadic conflict communication in a sample of 114 severe motor vehicle accident (MVA) survivors in a committed intimate relationship at the time of the accident. PTSD symptom severity and dyadic conflict communication were assessed at 4 and 16weeks post-MVA, and prospective associations were examined using path analysis. Total PTSD symptom severity at 4weeks prospectively predicted greater dysfunctional communication at 16weeks post-MVA but not vice versa. Examination at the level of PTSD symptom clusters revealed that effortful avoidance at 4weeks prospectively predicted greater dysfunctional communication at 16weeks, whereas dysfunctional communication 4weeks after the MVA predicted more severe emotional numbing at 16weeks. Findings highlight the role of PTSD symptoms in contributing to dysfunctional communication and the importance of considering PTSD symptom clusters separately when investigating the dynamic interplay between PTSD symptoms and relationship functioning over time, particularly during the early posttrauma period. Clinical implications for the prevention of chronic PTSD and associated relationship problems are discussed.

  4. ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chatelard, P., E-mail: patrick.chatelard@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Reinke, N.; Arndt, S. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France); Eckel, J. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50677 Köln (Germany); Jacq, F.; Marchetto, C.; Mun, C.; Piar, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, B.250, Cadarache BP3 13115, Saint-Paul-lez-Durance, Cedex (France)

    2014-06-01

    The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

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

  6. Comparative Analysis of the Pattern of Severe Injury Due to Road Traffic Accidents in Children

    Directory of Open Access Journals (Sweden)

    Ye. A. Spiridonova

    2010-01-01

    Full Text Available Objective: to study stepwise differences in the severity and pattern of severe traumatic injuries due to road traffic accidents in patients during the qualified and specialized stages of medical care in the age groups of 1 month to 18 years in the Rostov Region. Material and methods. The 2004—2009 case reports were used to make a retrospective comparative assessment of the condition of victims with severe road traffic injury at care stages in 2 groups: 1 one-month- to 18-year-old children who had been primarily admitted to the qualified-stage intensive care units (n=61; 2 one-month-to 18-year-old patients from the intensive care unit of the Regional Children’s Hospital, referred from the qualified-stage intensive care units (n=133. Results. The number of specialized-stage children in grave and extremely extensive grave condition at the specialized stage was 14.7% more than that at the qualified stage (100 and 85.3%, espectively. Concomitant injury was encountered more frequently (by 13% in the specialized-stage patients (73.7 and 60.7%, respectively. The severity of road traffic injury was determined mainly by brain injuries at the qualified irnd specialized stages (96.7 and 96.1%, by skeletal injuries (11.6% more frequently and thoracic ones (9.8% more frequently at the specialized stage. The prevalence of abdominal injuries at the qualified stage was 9.8% higher. The pattern of brain injury in the specialized-stage patients showed a preponderance of brain contusion and epidural hematomas by 18.5 and 6.5%, respectively. Conclusion. Severe thoracic and brain injuries (craniocerebral injuries, brain contusion, and intracranial hematomas are an indication for patient referral to the specialized care stage in order to perform high-technological diagnostic and therapeutic methods. Key words: children, road traffic accidents, severe injury, medical care stages.

  7. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  8. Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

    Directory of Open Access Journals (Sweden)

    Sandro Paci

    2012-01-01

    Full Text Available The SARNET2 (severe accidents Research NETwork of Excellence project started in April 2009 for 4 years in the 7th Framework Programme (FP7 of the European Commission (EC, following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs. The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers interested in SA management procedures.

  9. Development of highly reliable power and communication system for essential instruments under severe accidents in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Bo Hwan; Jang, Gi Chan; Shin, Sung Min; Kang, Hyun Gook; Rim, Chun Taek [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Soo Ill [I and C Group, Korea Hydro and Nuclear Power Co., Ltd, Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to 627 .deg. C and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

  10. New Solutions For Increasing Environmental Protection During Severe Accidents At Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kulyukhin, Sergei A.; Mikheev, Nikolai B. [Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow (Russian Federation); Falkovskii, Leo N.; Reshetov, Leo A.; Zvetkova, Marianna Ya. [All-Russian Research Institute of Atomic Machine-Building, Moscow, Russia (Russian Federation); Yagodkin, Ivan V.; Osipov, Viktor P.; Skvortsov, Sergei S. [Institute of Physics and Power Engineering, Obninsk (Russian Federation); Berkovich, Viktor M.; Taranov, Gennadii S.; Grigor' ev, Mikhail M. [Institute ' Atomenergoproekt' , Moscow (Russian Federation); Meshkov, Vladimir M.; Noskov, Andrei A.; Mitrofanov, Mikhail I. [ROSENERGOATOM Concern, Moscow (Russian Federation)

    2008-07-01

    This paper reports new solutions for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the inter-containment area, which includes the 'wet' stage (scrubbers, etc.), the 'dry' stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable. (authors)

  11. The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Jean-Pierre Van Dorsselaere

    2012-01-01

    Full Text Available Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP. After a first project in the 6th Framework Programme (FP6 of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…, source term issues (mainly iodine behaviour. The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

  12. In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Cronenberg, A.W. (Science and Engineering Associates, Inc., Albuquerque, NM (USA))

    1990-09-01

    In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

  13. Phenomenological and mechanistic modeling of melt-structure-water interactions in a light water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bui, V.A

    1998-10-01

    The objective of this work is to address the modeling of the thermal hydrodynamic phenomena and interactions occurring during the progression of reactor severe accidents. Integrated phenomenological models are developed to describe the accident scenarios, which consist of many processes, while mechanistic modeling, including direct numerical simulation, is carried out to describe separate effects and selected physical phenomena of particular importance 88 refs, 54 figs, 7 tabs

  14. 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...

  15. Research and development with regard to severe accidents in pressurised water reactors: Summary and outlook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This document reviews the current state of research on severe accidents in France and other countries. It aims to provide an objective vision, and one that's as exhaustive as possible, for this innovative field of research. It will help in identifying R and D requirements and categorising them hierarchically. Obviously, the resulting prioritisation must be completed by a rigorous examination of needs in terms of safety analyses for various risks and physical phenomena, especially in relation to Level 2 Probabilistic Safety Assessments. PSA-2 should be sufficiently advanced so as not to obscure physical phenomena that, if not properly understood, might result in substantial uncertainty. It should be noted that neither the safety analyses nor PSA-2 are presented in this document. This report describes the physical phenomena liable to occur during a severe accident, in the reactor vessel and the containment. It presents accident sequences and methods for limiting impact. The corresponding scenarios are detailed in Chapter 2. Chapter 3 deals with in-vessel accident progression, examining core degradation (3.1), corium behaviour in the lower head (3.2), vessel rupture (3.3) and high-pressure core meltdown (3.4). Chapter 4 focuses on phenomena liable to induce early containment failure, namely direct containment heating (4.1), hydrogen risk (4.2) and steam explosions (4.3). The phenomenon that could lead to a late containment failure, namely molten core-concrete interaction, is discussed in Chapter 5. Chapter 6 focuses on problems related to in-vessel and ex-vessel corium retention and cooling, namely in-vessel retention by flooding the primary circuit or the reactor pit (6.1), cooling of the corium under water during the corium-concrete interaction (6.2), corium spreading (6.3) and ex-vessel core catchers (6.4). Chapter 7 relates to the release and transport of fission products (FP), addressing the themes of in-vessel FP release (7.1) and ex-vessel FP release (7

  16. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  17. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

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

    Science.gov (United States)

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

    2014-10-01

    We conducted a latent class analysis (LCA) on 249 recent motor vehicle accident (MVA) victims to examine subgroups that differed in posttraumatic stress disorder (PTSD) symptom severity, current major depressive disorder and alcohol/other drug use disorders (MDD/AoDs), gender, and interpersonal trauma history 6-weeks post-MVA. A 4-class model best fit the data with a resilient class displaying asymptomatic PTSD symptom levels/low levels of comorbid disorders; a mild psychopathology class displaying mild PTSD symptom severity and current MDD; a moderate psychopathology class displaying severe PTSD symptom severity and current MDD/AoDs; and a severe psychopathology class displaying extreme PTSD symptom severity and current MDD. Classes also differed with respect to gender composition and history of interpersonal trauma experience. These findings may aid in the development of targeted interventions for recent MVA victims through the identification of subgroups distinguished by different patterns of psychiatric problems experienced 6-weeks post-MVA. Copyright © 2014 Elsevier Ltd. All rights reserved.

  19. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of a release parameters and metrorological data. According to the cost estimation for the basic scenarios, the population dependent cost is dominant. The cost for the protective actions such as evacuation and relocation have a small portion in the total cost and show little variation from scenario to scenario. The economic cost estimation for the seasonal scenarios show very similar trend as that for the basic scenarios. There are little or small variation in the economic cost for the different scenarios for each season except for the season-5 scenario. The health effect value shows maximum in Summer and minimum in Fall. On the contrast, the economic cost shows maximum in Fall and minimum in Summer. The result will be used as basic data in the establishment of effective emergency response and in the cost/benefit analysis in developing optimum risk reduction strategies.

  20. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  1. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of a release parameters and metrorological data. According to the cost estimation for the basic scenarios, the population dependent cost is dominant. The cost for the protective actions such as evacuation and relocation have a small portion in the total cost and show little variation from scenario to scenario. The economic cost estimation for the seasonal scenarios show very similar trend as that for the basic scenarios. There are little or small variation in the economic cost for the different scenarios for each season except for the season-5 scenario. The health effect value shows maximum in Summer and minimum in Fall. On the contrast, the economic cost shows maximum in Fall and minimum in Summer. The result will be used as basic data in the establishment of effective emergency response and in the cost/benefit analysis in developing optimum risk reduction strategies.

  2. 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.

  3. Ruthenium release modelling in air under severe accident conditions using the MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Beuzet, E.; Lamy, J.S. [EDF R and D, 1 avenue du General de Gaulle, F-92140 Clamart (France); Perron, H. [EDF R and D, Avenue des Renardieres, Ecuelles, F-77818 Moret sur Loing (France); Simoni, E. [Institut de Physique Nucleaire, Universite de Paris Sud XI, F-91406 Orsay (France)

    2010-07-01

    In a nuclear power plant (NPP), in some situations of low probability of severe accidents, an air ingress into the vessel occurs. Air is a highly oxidizing atmosphere that can lead to an enhanced core degradation affecting the release of Fission Products (FPs) to the environment (source term). Indeed, Zircaloy-4 cladding oxidation by air yields 85% more heat than by steam. Besides, UO{sub 2} can be oxidised to UO{sub 2+x} and mixed with Zr, which may lead to a decrease of the fuel melting temperature. Finally, air atmosphere can enhance the FPs release, noticeably that of ruthenium. Ruthenium is of particular interest for two main reasons: first, its high radiotoxicity due to its short and long half-life isotopes ({sup 103}Ru and {sup 106}Ru respectively) and second, its ability to form highly volatile compounds such as ruthenium gaseous tetra-oxide (RuO{sub 4}). Considering that the oxygen affinity decreases between cladding, fuel and ruthenium inclusions, it is of great need to understand the phenomena governing fuel oxidation by air and ruthenium release as prerequisites for the source term issues. A review of existing data on ruthenium release, controlled by fuel oxidation, leads us to implement a new model in the EDF version of MAAP4 severe accident code (Modular Accident Analysis Program). This model takes into account the fuel stoichiometric deviation and the oxygen partial pressure evolution inside the fuel to simulate its oxidation by air. Ruthenium is then oxidised. Its oxides are released by volatilisation above the fuel. All the different ruthenium oxides formed and released are taken into consideration in the model, in terms of their particular reaction constants. In this way, partial pressures of ruthenium oxides are given in the atmosphere so that it is possible to know the fraction of ruthenium released in the atmosphere. This new model has been assessed against an analytical test of FPs release in air atmosphere performed at CEA (VERCORS RT8). The

  4. A Study on Fission Product Behavior during a Severe Accident at APR1400 Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Song-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, calculations have been carried out for a SBO sequence similar to the selected scenario, but a faster one with simple assumptions. Instead, a sensitivity study was carried out to take into account the effects of such differences on the fission product behavior. Probabilistic Safety Assessment (PSA) for Shin-Kori 3·4 nuclear power plants, which are APR1400 type reactors, were reviewed. After all, the representing scenarios were determined to be the sequences with station blackout (SBO), interfacing system LOCA (ISLOCA), and steam generator tube rupture (SGTR), which are similar to those of the U.S.NRC's State-of-the-Art Reactor Consequence Analyses (SOARCA) study. Among those sequences, SBO occupies the largest portion of the risk from severe accidents, and was selected to be analyzed at first about the fission product behavior in the containment. It includes events such as failure of the alternative AC power generator following a blackout event, successful operation of turbine-driven auxiliary feed water (AFW) pump, late recovery of offsite power before containment failure, in-vessel injection and successful actuation of cavity flooding system and spray system, and failure of hydrogen mitigation system. We use MELCOR 1.8.6 with the 35- and 2-cell compartment models of the containment. Since MELCOR does not treat organic iodide, we tried to make the results up by MELCOR-RAIM which is the MELCOR code coupled with RAIM, a stand-alone code developed for evaluation of the iodine behavior. In order to investigate the fission product behavior during a severe accident at APR1400, we have selected the representing scenarios with SBO, ISLOCA and SGTR. Among them, a SBO sequence similar to the selected scenario, but a faster one with simple assumptions, was analyzed using MELCOR v1.8.6 with 35-cell models of the containment. For the sensitivity analysis, we use the 2-cell containment model and the codes with the iodine chemistry model such as MELCOR with

  5. Study on corium behavior in the reactor cavity during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The report contains following four results of studies on molten corium-concrete interaction, which has been recognized as important aspects of severe reactor accident; 1. MELCOR code modification has been performed for heat transfer model between ex-vessel molten corium and overlying water pool. The existing model do not consider debris particulation and water penetration in the ex-vessel debris cooling. The new model employs dryout heat flux in determining the heat removal from a debris bed by water penetration. 2. A parametric model which can evaluate ex-vessel concrete erosions has been developed. The model is expected to evaluate the concrete erosion in a limited error range with only a little effort. The model has been derived by the sensitivity studies using MELCOR and MAAP programs. 3. During the corium-concrete interaction, there is a temperature distribution inside basemat concrete. MELCOR calculates concrete response based on one-dimensional steady-state ablation, with no consideration given to conduction into the concrete or to decomposition in advance of the ablation front. Thus there is a necessity to improve the concrete decomposition model in MELCOR. In this report the transient conduction model and the methodology of implementation into MELCOR were suggested. 4. Major modeling assumptions and limits of MELTSPREAD-1, which is a transient one-dimensional computer code to predict the gravity-driven spreading of molten corium in the reactor cavity under severe accidents, are evaluated via review of general conservation equations and used models. The models being reviewed include heat transfer models at melt lower/upper surfaces, a concrete dryout model, and a shell heatup model. The evaluation results suggest the degree of MELTSPREAD-1 approximation compared with real spreading flow and the strong/weak points or restrictions of the code. 17 refs., 19 figs., 6 tabs. (Author)

  6. The influence of the crust layer on RPV structural failure under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Jianfeng, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Li, Xiangqing [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Bao, Shiyi [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, Lijia [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Gao, Zengliang [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2017-05-15

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  7. Prediction of rate and severity of adverse perioperative outcomes: "normal accidents" revisited.

    Science.gov (United States)

    Saubermann, Albert J; Lagasse, Robert S

    2012-01-01

    The American Society of Anesthesiologists Physical Status classification system has been shown to predict the frequency of perioperative morbidity and mortality despite known subjectivity, inconsistent application, and exclusion of many perioperative confounding variables. The authors examined the relationship between the American Society of Anesthesiologists Physical Status and both the frequency and the severity of adverse events over a 10-year period in an academic anesthesiology practice. The American Society of Anesthesiologists Physical Status is predictive of not only the frequency of adverse perioperative events, but also the severity of adverse events. These nonlinear mathematical relationships can provide meaningful information on performance and risk. Calculated odds ratios allow discussion about individualized anesthesia risks based on the American Society of Anesthesiologists Physical Status because the added complexity of the surgical or diagnostic procedure, and other perioperative confounding variables, is indirectly factored into the Physical Status classification. The ability of the American Society of Anesthesiologists Physical Status to predict adverse outcome frequency and severity in a nonlinear relationship can be fully explained by applying the Normal Accident Theory, a well-known theory of system failure that relates the interactive complexity of system components to the frequency and the severity of system failures or adverse events.

  8. Effective Factors in Severity of Traffic Accident-Related Traumas; an Epidemiologic Study Based on the Haddon Matrix

    Directory of Open Access Journals (Sweden)

    Kambiz Masoumi

    2016-04-01

    Full Text Available Introduction: Traffic accidents are the 8th cause of mortality in different countries and are expected to rise to the 3rd rank by 2020. Based on the Haddon matrix numerous factors such as environment, host, and agent can affect the severity of traffic-related traumas. Therefore, the present study aimed to evaluate the effective factors in severity of these traumas based on Haddon matrix. Methods: In the present 1-month cross-sectional study, all the patients injured in traffic accidents, who were referred to the ED of Imam Khomeini and Golestan Hospitals, Ahvaz, Iran, during March 2013 were evaluated. Based on the Haddon matrix, effective factors in accident occurrence were defined in 3 groups of host, agent, and environment. Demographic data of the patients and data regarding Haddon risk factors were extracted and analyzed using SPSS version 20. Results: 700 injured people with the mean age of 29.66 ± 12.64 years (3-82 were evaluated (92.4% male. Trauma mechanism was car-pedestrian in 308 (44% of the cases and car-motorcycle in 175 (25%. 610 (87.1% cases were traffic accidents and 371 (53% occurred in the time between 2 pm and 8 pm. Violation of speed limit was the most common violation with 570 (81.4% cases, followed by violation of right-of-way in 57 (8.1% patients. 59.9% of the severe and critical injuries had occurred on road accidents, while 61.3% of the injuries caused by traffic accidents were mild to moderate (p < 0.001. The most common mechanisms of trauma for critical injuries were rollover (72.5%, motorcycle-pedestrian (23.8%, and car-motorcycle (13.14% accidents (p < 0.001. Conclusion: Based on the results of the present study, the most important effective factors in severity of traffic accident-related traumas were age over 50, not using safety tools, and undertaking among host-related factors; insufficient environment safety, road accidents and time between 2 pm and 8 pm among environmental factors; and finally, rollover, car

  9. Management of a severe accident on a pressurised water reactor in France; La gestion d'un accident grave sur un reacteur a eau sous pression en France

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This brief document defines what a severe accident is on a nuclear reactor, indicates the different failure modes which have been defined (vapour explosion in the reactor vessel, hydrogen explosion, and so on). It describes the management of a core fusion accident for pressurized water reactors, for which a guide has been designed, the GIAG (intervention guide for a severe accident situation). The principles of such an intervention are described, and then the approach for an EPR reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-05-01

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

  11. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

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

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-11-01

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

  14. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  15. 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.

  16. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V.; Rosenberg, R. [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  17. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Daejeon (Korea, Republic of)

    2015-10-15

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  18. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  19. Making the journey safe: recognising and responding to severe sepsis in accident and emergency

    Science.gov (United States)

    Pinnington, Sarah; Atterton, Brigid; Ingleby, Sarah

    2016-01-01

    Severe sepsis is a clinical emergency. Despite the nationwide recognition of the sepsis six treatment bundle as the first line emergency treatment for this presentation, compliance in sepsis six provision remains inadequately low. The project goals were to improve compliance with the implementation of the Sepsis Six in patients with severe sepsis and/or septic shock. In improving timely care delivery it was anticipated improvements would be made in relation to patient safety and experience, and reductions in length of stay (LoS) and mortality. The project intended to make the pathway for those presenting with sepsis safe and consistent, where sepsis is recognised and treated in a timely manner according to best practice. The aim of the project was to understand the what the barriers where to providing safe effective care for the patient presenting with severe sepsis in A&E. Using the Safer Clinical Systems (SCS) tools developed byte Health Foundation and Warwick University, the project team identified the hazards and associated risks in the septic patient pathway. The level of analysis employed enabled the project team to identify the major risks, themes, and factors of influence within this pathway. The analysis identified twenty nine possible interventions, of which six were chosen following option appraisal. Further interventions were recommended to the accident and emergency as part of a business case and further changes in process. Audits identified all severely septic patients presenting to A&E in October 2014 (n=67) and post intervention in September 2015 (n=93). Compared analysis demonstrated an increase in compliance with the implementation of the sepsis six care bundle from 7% to 41%, a reduction in LoS by 1.9 days and a decrease in 30 day mortality by 50%. Additional audit reviewed the management of 10 septic patients per week for the duration of the project to assess the real time impact of the selected interventions.

  20. Phenomenology of severe accidents in BWR type reactors. First part; Fenomenologia de accidentes severos en reactores nucleares de agua en ebullicion. Primera parte

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [Instituto de Investigaciones Electricas, Gerencia de Energia Nuclear, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)

    2003-07-01

    A Severe Accident in a nuclear power plant is a deviation from its normal operating conditions, resulting in substantial damage to the core and, potentially, the release of fission products. Although the occurrence of a Severe Accident on a nuclear power plant is a low probability event, due to the multiple safety systems and strict safety regulations applied since plant design and during operation, Severe Accident Analysis is performed as a safety proactive activity. Nuclear Power Plant Severe Accident Analysis is of great benefit for safety studies, training and accident management, among other applications. This work describes and summarizes some of the most important phenomena in Severe Accident field and briefly illustrates its potential use based on the results of two generic simulations. Equally important and abundant as those here presented, fission product transport and retention phenomena are deferred to a complementary work. (Author)

  1. Man, road and vehicle: risk factors associated with the severity of traffic accidents.

    Science.gov (United States)

    Almeida, Rosa Lívia Freitas de; Bezerra Filho, José Gomes; Braga, José Ueleres; Magalhães, Francismeire Brasileiro; Macedo, Marinila Calderaro Munguba; Silva, Kellyanne Abreu

    2013-08-01

    To describe the main characteristics of victims, roads and vehicles involved in traffic accidents and the risk factors involved in accidents resulting in death. METHODS A non-concurrent cohort study of traffic accidents in Fortaleza, CE, Northeastern Brazil, in the period from January 2004 to December 2008. Data from the Fortaleza Traffic Accidents Information System, the Mortality Information System, the Hospital Information System and the State Traffic Department Driving Licenses and Vehicle database. Deterministic and probabilistic relationship techniques were used to integrate the databases. First, descriptive analysis of data relating to people, roads, vehicles and weather was carried out. In the investigation of risk factors for death by traffic accident, generalized linear models were used. The fit of the model was verified by likelihood ratio and ROC analysis. RESULTS There were 118,830 accidents recorded in the period. The most common types of accidents were crashes/collisions (78.1%), running over pedestrians (11.9%), colliding with a fixed obstacle (3.9%), and with motorcycles (18.1%). Deaths occurred in 1.4% of accidents. The factors that were independently associated with death by traffic accident in the final model were bicycles (OR = 21.2, 95%CI 16.1;27.8), running over pedestrians OR = 5.9 (95%CI 3.7;9.2), collision with a fixed obstacle (OR = 5.7, 95%CI 3.1;10.5) and accidents involving motorcyclists (OR = 3.5, 95%CI 2.6;4.6). The main contributing factors were a single person being involved (OR = 6.6, 95%CI 4.1;10.73), presence of unskilled drivers (OR = 4.1, 95%CI 2.9;5.5) a single vehicle (OR = 3.9, 95%CI 2,3;6,4), male (OR = 2.5, 95%CI 1.9;3.3), traffic on roads under federal jurisdiction (OR = 2.4, 95%CI 1.8;3.7), early morning hours (OR = 2.4, 95%CI 1.8;3.0), and Sundays (OR = 1.7, 95%CI 1.3;2.2), adjusted according to the log-binomial model. CONCLUSIONS Activities promoting the prevention of traffic accidents should primarily focus on

  2. RAIM-A model for iodine behavior in containment under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Chul; Cho, Yeong Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-12-15

    Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  3. Measurement of buckling load for metallic plate columns in severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Byeongnam, E-mail: jo@vis.t.u-tokyo.ac.jp; Sagawa, Wataru, E-mail: sagawa@vis.t.u-tokyo.ac.jp; Okamoto, Koji, E-mail: okamoto@n.t.u-tokyo.ac.jp

    2014-07-01

    Highlights: • Buckling load was experimentally measured in a wide range of temperature up to 1200 °C. • Two different test methods for measuring buckling failure load were suggested and compared. • Creep buckling under compressive load was performed to explain results of buckling tests. • Reduced buckling load was explained by effects of creep buckling, geometrical imperfection, and thermal stress. • Buckling processes were visualized by a high speed camera. - Abstract: In severe accidents, a reactor pressure vessel, its components, and piping have to be under extremely high temperature and high pressure conditions, which results in failure modes like rupture by internal pressure, buckling, creep, and their combinations. In this study, buckling (failure) load was experimentally measured for metallic columns under the compressive force from room temperature up to 1200 °C. A stainless steel was chosen to be a test material to measure the buckling load. Two different test methods were employed to explore the effect of thermal history of the material on the buckling load. Particularly, the effect of creep under a compressive load was considered as a reason for the reduced buckling load at high temperatures. Additionally, finite element simulations were also conducted to predict buckling load for both an ideal column and a column with geometrical imperfection as well. Moreover, buckling process was visualized using a high speed camera to understand buckling processes.

  4. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  5. Reclamation of contaminated urban and rural environments following a severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Strand, P.; Skuterud, L. [eds.] [Norwegian Radiation Protection Authority (Norway); Melin, J. [ed.] [Swedish Radiation Protection Institute (Sweden)

    1997-10-01

    In the event of a severe nuclear accident releasing radioactive materials to the atmosphere, there is a potential for widespread contamination of both the urban and rural environments. In some instances of environmental contamination, natural processes may eventually reduce or eliminate the problem without man`s intervention. The situation with respect to radioactive contamination is no different except that radioactive contamination will also disappear through normal physical radioactive decay. In other cases, man is often able to mitigate potential harmful effects by cleaning, washing, abrading or by the application of chemicals. The actions taken by man to mitigate the potential harmful effects of contamination are described as countermeasures. In the case of radioactive contamination, the objective of countermeasures is to minimise radiation doses to man. This document is intended as a guide to those groups who may, at very short notice, be called upon to manage and reclaim radioactively contaminated urban and rural environments in the Nordic countries. However, much of the information and recommendations are also equally applicable in other countries. The document is divided into eight distinct parts, namely: 1. The Urban Environment; 2. The Cultivated Agricultural Environment; 3. Animals; 4. Forests; 5. Freshwater and Fish; 6. Management and Disposal of Radioactive Waste from Clean-up Operations; 7. Radiation Protection and Safety of Clean-up Operators; 8. Resources Available in Society. (EG).

  6. The severe accident research programme PHEBUS F.P.: First results and future tests

    Energy Technology Data Exchange (ETDEWEB)

    Schwarz, M. [Institut de Protection et de Surete Nucleaire IPSN, Saint Paul Lez Durance (France); Hardt, P. von der [Joint Research Centre - Safety Technology Institute, Saint Paul Lez Durance (France)

    1996-03-01

    PHEBUS FP is an international programme, managed by the French Institut de Protection et de Surete Nucleaire, Electricite de France and the European Commission in close collaboration with the USNRC (US), COG (Canada), NUPEC and JAERI (Japan) and KAERI (South Korea). Its objective is to investigate through a series of in-pile integral experiments, key phenomena involved in LWR severe accident such as the degradation of core materials up to molten pool, the subsequent release of fission products and of structural materials, their transport in the cooling system and their deposition in the containment with a special emphasis on the volatility of iodine. After a general programme description, the paper focuses on the status of analysis of the first test FPT-0, which involved trace irradiated fuel and which has shown some quite unexpected results regarding fuel degradation and iodine behaviour, and on the upcoming test FPT-1 which will use irradiated fuel. The status of the preparation of the remaining tests of the programme is also presented.

  7. The Need to introduce CFD Methodology in Analyze Hydrogen Distribution for Postulated Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Na, Hanbee; Park, Sukyung; Kim, Kyuntae [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Jongkwang [Hanbat National University, Daejeon (Korea, Republic of); Kwon, Sejin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    The regulatory requirements for combustible gas control systems in Korea is that mean hydrogen mole fraction shall be lower than 10 %, containment integrity shall be kept from combustion of hydrogen, and detonation and global fast turbulent combustion shall be avoided. KHNP provided some analysis which show hydrogen mole fraction is less than 10 % and detonation and global fast turbulence combustion are avoided for postulated severe accident events which covered over 90 % of CDF (core damage frequency) for each NPP. The results were from MAAP code that can simulate from the initiation of the accidents to hydrogen distribution inside containments. It is a Lumped-Parameter codes in which the transport of energy and mass is possible in only predetermined one direction. Therefore, there has been a long-history dispute whether one-dimensional LP codes could simulate the transportation of hydrogen accurately. For example, KHNP made a MAAP model to simulate hydrogen distribution in KSNP (Korean Standard Nuclear Plants), and the containment free volume is divided into 27 nodes in which it is assumed all the properties like each molecule mole fraction and temperate are uniform in each node. In addition, the maximum volume size of them is over 22,000 m{sup 3}, and it is not quite confident that the mole fraction of each molecules and temperature are uniform in the big size space. As for the stress test results of the Wolsong 1, civil experts asked KHNP to conduct hydrogen distribution analysis using Computational Fluid Dynamics (CFD) methodology, and if needed to install hydrogen ignitors in Wolsong 1 NPP. As a reviewer for KHNP's post actions to the Stress Test, the author also asked KHNP to do CFD analysis of hydrogen distribution, and KHNP finally agreed to analyze it using CFD by 2017. KHNP submitted a Shin-hanul 1 and 2 Operation License application in 2015, and the author also asked it to do CFD analysis to simulate hydrogen distribution for Shin-hanul 1 and 2

  8. Assessment of Spatial Unevenness of Road Accidents Severity as Instrument of Preventive Protection from Emergency Situations in Road Complex

    Science.gov (United States)

    Petrov, A.; Petrova, D.

    2016-08-01

    Emergency situations in road complex are road traffic accidents (RA) with severe consequences. These are incidents connected with the death and injury of large number of people. The most common reasons for this are the collision of three or more cars, the collision of buses with trains at railroad crossings, the fall of the buses in the mountain gorge, and other similar cases. Is it possible to predict such events? How to build a preventive protection against such emergencies? We have to understand that emergencies in a road complex are qualitative expression of the quantitative processes that characterize the general state of road safety in the region. In this regard, at the level of state monitoring of emergency situations it is important to understand in general - in which region the situation is more complicated and in which is more favorable. This knowledge helps to more efficiently reallocate resources intended to solve the problems of road safety provision. The consequence of this is improvement of the quality of preventive protection from the emergencies in the road complex. The article presents quantitative values of severity of accidents in the Russian Federation regions and the Pareto chart distribution of cumulates of the accident severity for the Russian Federation. On the basis of the complex assessment of the spatial non-uniformity of the accident severity results it offers two important recommendations, implementation of which will alleviate the issue of formation of emergency situations in the road of the Russian Federation on the basis of the complex assessment of the spatial nonuniformity of the accident severity results.

  9. Assessing injury severity in bicyclists involved in traffic accidents to more effectively prevent fatal bicycle injuries in Japan.

    Science.gov (United States)

    Gomei, Sayaka; Hitosugi, Masahito; Ikegami, Keiichi; Tokudome, Shogo

    2013-10-01

    The objective of this study was to clarify the relationship between injury severity in bicyclists involved in traffic accidents and patient outcome or type of vehicle involved in order to propose effective measures to prevent fatal bicycle injuries. Hospital records were reviewed for all patients from 2007 to 2010 who had been involved in a traffic accident while riding a bicycle and were subsequently transferred to the Shock Trauma Center of Dokkyo Medical University Koshigaya Hospital. Patient outcomes and type of vehicle that caused the injury were examined. The mechanism of injury, Abbreviated Injury Scale (AIS) score, and Injury Severity Score (ISS) of the patient were determined. A total of 115 patients' records were reviewed. The mean patient age was 47.1 ± 27.4 years. The average ISS was 23.9, with an average maximum AIS (MAIS) score of 3.7. The ISS, MAIS score, head AIS score, and chest AIS score were well correlated with patient outcome. The head AIS score was significantly higher in patients who had died (mean of 4.4); however, the ISS, MAIS score, and head AIS score did not differ significantly according to the type of vehicle involved in the accident. The mean head AIS scores were as high as 2.4 or more for accidents involving any type of vehicle. This study provides useful information for forensic pathologists who suspect head injuries in bicyclists involved in traffic accidents. To effectively reduce bicyclist fatalities from traffic accidents, helmet use should be required for all bicyclists.

  10. 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.

  11. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  12. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.D.; Kmetyk, L.N.; Whitehead, D.; Miller, L. [Sandia National Labs., Albuquerque, NM (United States); Forester, J. [Science Applications International Corp., Albuquerque, NM (United States); Johnson, J. [GRAM, Inc., Albuquerque, NM (United States)

    1995-03-01

    Traditionally, probabilistic risk assessments (PRAS) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Recent studies and operational experience have, however, implied that accidents during low power and shutdown could be significant contributors to risk. In response to this concern, in 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The program consists of two parallel projects being performed by Brookhaven National Laboratory (Surry) and Sandia National Laboratories (Grand Gulf). The program objectives include assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing the estimated risks with the risk associated with accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program is that of a Level-3 PRA. The subject of this report is the PRA of the Grand Gulf Nuclear Station, Unit 1. The Grand Gulf plant utilizes a 3833 MWt BUR-6 boiling water reactor housed in a Mark III containment. The Grand Gulf plant is located near Port Gibson, Mississippi. The regime of shutdown analyzed in this study was plant operational state (POS) 5 during a refueling outage, which is approximately Cold Shutdown as defined by Grand Gulf Technical Specifications. The entire PRA of POS 5 is documented in a multi-volume NUREG report (NUREG/CR-6143). The internal events accident sequence analysis (Level 1) is documented in Volume 2. The Level 1 internal fire and internal flood analyses are documented in Vols 3 and 4, respectively.

  14. Coolability of corium debris under severe accident conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Saidur

    2013-11-15

    The debris bed which may be formed in different stages of a severe accident will be hot and heated by decay heat from the radioactive fission products. In order to establish a steady state of long-term cooling, this hot debris needs to be quenched at first. If quenching by water ingression into the dry bed is not rapid enough then heat-up by decay heat in still dry regions may again yield melting. Thus, chances of coolability must be investigated considering quenching against heat-up due to decay heat, in the context of reactor safety research. As a basis of the present investigations, models for simulation of two phase flow through porous medium were already available in the MEWA code, being under development at IKE. The objective of this thesis is to apply the code in essential phases of severe accidents and to investigate the chances, options and measures for coolability. Further, within the tasks, improvements to remove weaknesses in modeling and implementation of extensions concerning missing parts are included. It was identified previously that classical models without explicit considering the interfacial friction, can predict dryout heat flux (DHF) well under top fed condition but under-predict DHF values under bottom flooding conditions. Tung and Dhir introduced an interfacial friction term in their model, but this model has deficits for smaller particles considered as relevant for reactor conditions. Therefore, some modification of Tung and Dhir model is proposed in the present work to extent it for smaller particles. A significant improvement with the new friction description (Modified Tung and Dhir, MTD) is obtained considering the aim of a unified description for both top and bottom flooding conditions and for broad bandwidth of bed conditions. Calculations for reactor conditions are carried out in order to explore whether or to which degree coolability can be concluded, how strong the trend to coolability is and where major limits occur. The general

  15. Experiments and analyses on melt-structure-water interactions during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Seghal, B.R.; Dinh, T.N.; Bui, V.A.; Green, J.A.; Nourgaliev, R.R.; Okkonen, T.O.; Dinh, A.T. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-04-01

    This report is the final report for the research project Melt Structure Water Interactions (MSWI). It describes results of analytical and experimental studies concerning MSWI during the course of a hypothetical core meltdown accident in a LWR. Emphasis has been placed on phenomena which govern vessel failure mode and timing and the mechanisms and properties which govern the fragmentation and breakup of melt jets and droplets. It was found that: 2-D effects significantly diminished the focusing effect of an overlying metallic layer on top of an oxide melt pool. This result improves the feasibility of in-vessel retention of a melt pool through external cooling of the lower head; phenomena related to hole ablation and melt discharge, in the event of vessel failure, are affected significantly by crust formation; the jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters but also on the melt physical properties, which change as the melt cools down from liquid to solid temperature; film boiling was investigated by developing a two-phase flow model and inserting it in a multi-D fluid dynamics code. It was concluded that the thickness of the film on the surface of a melt jet would be small and that the effects of the film on the process should not be large. This conclusion is contrary to the modeling employed in some other codes. The computer codes were developed and validated against the data obtained in the MSWI Project. The melt vessel interaction thermal analysis code describes the process of melt pool formation and convection and the resulting vessel thermal loadings. In addition, several innovative models were developed to describe the melt-water interaction process. The code MELT-3D treats the melt jet as a collection of particles whose movement is described with a three-dimensional Eulerian formulation. The model (SIPHRA) tracks the melt jet with an additional equation, using the

  16. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  17. Iodine chemistry at severe accidents. A review and evaluation of the state-of-the-art in the field. APRI 5 report. Part I: Iodine chemistry at hypothetical severe accidents. A review of the state-of-the-art 2003. Part II: A comparison of our knowledge on iodine chemistry and fission products with the current models used in MAAP 4.0.5; Jodkemi under svaara haverier. En sammanstaellnig och vaerdering av kunskapslaeget inom omraadet. APRI 5 rapport. Del I: Jodkemi vid hypotetiska svaara haverier. En genomgaang av kunskapslaeget aar 2003. Del II: Jaemfoerelse av kunskapslaeget om jodkemi och fissionsprodukter med aktuella modeller i MAAP 4.0.5

    Energy Technology Data Exchange (ETDEWEB)

    Liljenzin, Jan-Olov [Liljenzins data och kemikonsult, Goeteborg (Sweden)

    2005-01-01

    The current report tries to summarize and analyze the state-of-the-art on Iodine chemistry relevant to the conditions expected during severe accidents in nuclear power plants. This has made it necessary to compare a considerable amount of data, new as well as old, in order to try to find the reasons behind some changes in the expected chemical behaviour of Iodine. In a few cases this has been far from simple. Many numerical values are given in this report. However, me numbers given should not be used in a non-critical way because they are often deduced from measurements whose interpretation depends on various kinds of systematic differences and assumptions with regard to technique, 'known' constants, and models applied. The most important observation today is that one can no longer uncritically assume that iodine is only released and transported as cesium iodide. The considerable effect that control rod material (including other construction materials) can have on the way in which an accident develops and on its iodine chemistry is clearly seen from the results of the experiments performed within the PHEBUS FP project. The second part of the report evaluates new knowledge on Iodine chemistry and Iodine behaviour of importance in severe nuclear reactor accidents. Also some new information regarding the behaviour and chemistry of other fission products has been collected. In the light of this information, the current modelling of Iodine behaviour in the MAAP code version 4.0.5 has been investigated. No modelling errors have been found. However, some of the equations used to calculate the vapour pressure of the components in the AlC-alloy used in PWR control rods give questionable results. An error in the MAAP manual was found which should be corrected. Finally, some suggestions are given for future improvements in the modelling of severe accidents used in MAAP for both BWRs and PWRs.

  18. An Entry Point of the Emergency Response Robot for Management of Severe Accident of the Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaiwan; Jeong, Kyungmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this paper, from the view point of DID (defense-in depth), we discuss the entry point of the nuclear emergency response robot to cope with a nuclear disaster. A Japanese nuclear disaster preparedness robot system was developed, after the JCO criticality accident in 1999, to cope with INES (International Nuclear and Radiological Event Scale) Level 3 serious incidents. INES Level 3 means the loss of DID (defense-in-depth) functions. It also indicates that ESF (engineered safety features) and ECCS (emergency core cooling system) resources, which are used to prevent serious incidents from escalating to severe accidents (core melt-down), have been almost exhausted. In the unit 1 reactor accident of Fukushima Daiichi Nuclear Power Plant, escalation from INES Level 1 (Out of Limiting Condition for Operation) to INES Level 5 (serious core melting-down) took less than two hours. Major facts are briefly described here in based on data gathered immediately after the tsunami over Fukushima Daiichi Nuclear Power Plant. Ο 15:35 on March 11, 2nd tsunami arrived. - 15:37, SBO (station black out) Ο 15:42, Interprets as a SBO (INES Level 1) - Loss of DC power for Instrumentation (Unknown of reactor water level) Ο 16:36, Loss of ECCS function (INELS Level 5) (Entry into a BDBA status) The Moni ROBO-A robot of the Japan Nuclear Safety Technology Center (NUSTEC) was a nuclear disaster preparedness robot developed after the JCO criticality accident. It was the only robot that had been steadily maintained and was available at the time of the Fukushima Daiichi Nuclear Power Plant accident. However, it was not helpful in mitigating the accident because it is assumed to have arrived at J-Village after the accident had been escalated to INES Level 5 or higher. Based on the paper by S. Kawatsuma of JAEA and response data gathered immediately after the tsunami, it is estimated that the NUSTEC's Moni ROBO-A arrived at J-Village after the designed entry point for INES Level 3

  19. 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

  20. Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Haraldsson, H.O.; Li, H.X.; Konovakhin, M.; Paladino, D.; Leung, W.H [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1999-08-01

    This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity

  1. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Jo, J.; Lin, C.C.; Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1995-05-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. 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 results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6.

  2. Evaluation of severe accident risks and the potential for risk reduction: Surry Power Station, Unit 1: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, A.S.; Boyd, G.J.; Kunsman, D.M.; Murfin, W.B.; Williams, D.C.

    1987-02-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a particular pressurized water reactor with a subatmospheric containment (Surry, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally lower than previously evaluated in the Reactor Safety Study (RSS). However, certain unresolved issues (such as direct containment heating) caused the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. None of the postulated safety options appears to be cost effective for the Surry power plant. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150.

  3. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  4. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  5. Phenomenological studies on melt-structure-water interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Haraldsson, H.O.; Nourgaliev, R.R.; Konovalikhin, M.; Paladino, D.; Gubaidullin, A.A.; Kolb, G.; Theerthan, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    2000-05-01

    This is the annual report for the work performed in 1999 in the research project Melt-Structure-Water Interactions During Severe Accidents in LWRs, under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The emphasis of the work is placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. We believe that significant technical advances have been achieved during the course of these studies. It was found that: The coolant temperature has significant influence on the characteristics of debris fragments produced from the breakup of an oxidic melt jet. At low subcooling the fragments are relatively large and irregular compared to the smaller particles produced at high subcooling. The melt jet density has considerable effect on the fragment size produced. As the melt density increases the fragment size becomes smaller. The mass mean size of the debris changes proportionally to the square root of the coolant to melt density ratio. The melt superheat has little effect on the debris particle size distribution produced during the melt jet fragmentation. The impingement velocity of the jet has significant impact on the fragmentation process. At lower jet velocity the melt fragments agglomerate and form a cake of large size debris. When the jet velocity is increased more complete fragmentation is obtained. The scaling methodology for melt spreading, developed during 1998, has been further validated against almost all of the spreading experimental data available so far. Experimental results for the dryout heat flux of homogeneous particulate debris beds with top flooding compare well with the Lipinski correlation. For the stratified particle beds, the fine particle layer resting on the top of another particle layer dominates the dryout processes

  6. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  7. Criticality accident in uranium fuel processing plant. Emergency medical care and dose estimation for the severely overexposed patients

    Energy Technology Data Exchange (ETDEWEB)

    Akashi, Makoto; Ishigure, Nobuhito [National Inst. of Radiological Sciences, Chiba (Japan)

    2000-08-01

    A criticality accident occurred in JCO, a plant for nuclear fuel production in 1999 and three workers were exposed to extremely high-level radiation (neutron and {gamma}-ray). This report describes outlines of the clinical courses and the medical cares for the patients of this accident and the emergent medical system for radiation accident in Japan. One (A) of the three workers of JCO had vomiting and diarrhea within several minutes after the accident and another one (B) had also vomiting within one hour after. Based on these evidences, the exposure dose of A and B were estimated to be more than 8 and 4 GyEq, respectively. Generally, acute radiation syndrome (ARS) is assigned into three phases; prodromal phase, critical or manifestation phase and recovery phase or death. In the prodromal phase, anorexia, nausea, vomiting and diarrhea often develop, whereas the second phase is asymptotic. In the third phase, various syndromes including infection, hemorrhage, dehydration shock and neurotic syndromes are apt to occur. It is known that radiation exposure at 1 Gy or more might induce such acute radiation syndromes. Based on the clinical findings of Chernobyl accident, it has been thought that exposure at 0.5 Gy or more causes a lowering of lymphocyte level and a decrease in immunological activities within 48 hours. Lymphocyte count is available as an indicator for the evaluation of exposure dose in early phase, but not in later phase The three workers of JCO underwent chemical analysis of blood components, chromosomal analysis and analysis of blood {sup 24}Na immediately after the arrival at National Institute of Radiological Sciences via National Mito Hospital specified as the third and the second facility for the emergency medical care system in Japan, respectively. (M.N.)

  8. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  9. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  10. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration.

  11. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration.

  12. 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.

  13. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  14. Studies of the UO 2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

    Science.gov (United States)

    Shiozawa, S.; Ichikawa, M.; Fujishiro, T.

    1988-06-01

    Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO 2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

  15. 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.

  16. 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.

  17. A Study on Licensing Requirement for Severe Accident of PGSFR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kilyoo; Han, Sang Hoon; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Metal-fueled SFRs such as PRISM, 4S, etc. do not have serious SA research since they are inherently safe with low melting point, high boiling temperature, good thermal inertia due to good conductivity, and passive safety systems, etc. Since PGSFR is one of the metal-fueled SFRs which have been developed to overcome the SAs drawbacks of oxide fueled SFRs, there would occur no SAs in view of the then-current criteria of SAs. Unfortunately, there is no SA regulatory requirement for SFR in Korea and in USA. Thus, we can think the following two options to get the design approval for PGSFR. Option 1: Methods and strategies to prevent and mitigate SAs of PGSFR should be reported, and for which required research should be performed. After Fukushima accident, it seems that this SA perspective becomes important. Option 2: Since PGSFR or a metal-fueled SFR such as PRISM is inherently too safe, there is no SA as proved by the EBR II experiment. Thus, the issues of SAs were already solved, and SA research is not necessary. In this paper, by reviewing of the recent nuclear regulation trend in Korean and in USA, and by checking of PGSFR PSA model, which option is better in the design approval for PGSFR is discussed. Although the inherent and passive safety measures of PGSFR could satisfy with the then-current regulatory requirement used in the pre-application of PRISM, the trend in U. S and Korean nuclear regulatory after Fukushima accident shows that SA cannot be treated in residual risk category. Rather, after setting SA scenario, further SA research should be done which has not been well performed after PRISM pre-application in 1994. Especially, PGSFR should cope with the extended SBO requirement and triple failures issued in Fukushima accident. Although accurate SA scenarios for PGSFR would be identified after performing the external PSA for PGSFR, some triple faults are suggested as SA scenarios.

  18. Work Incapacity and Treatment Costs After Severe Accidents: Standard Versus Intensive Case Management in a 6-Year Randomized Controlled Trial.

    Science.gov (United States)

    Scholz, Stefan M; Andermatt, Peter; Tobler, Benno L; Spinnler, Dieter

    2016-09-01

    Purpose Case management is widely accepted as an effective method to support medical rehabilitation and vocational reintegration of accident victims with musculoskeletal injuries. This study investigates whether more intensive case management improves outcomes such as work incapacity and treatment costs for severely injured patients. Methods 8,050 patients were randomly allocated either to standard case management (SCM, administered by claims specialists) or intensive case management (ICM, administered by case managers). These study groups differ mainly by caseload, which was approximately 100 cases in SCM and 35 in ICM. The setting is equivalent to a prospective randomized controlled trial. A 6-year follow-up period was chosen in order to encompass both short-term insurance benefits and permanent disability costs. All data were extracted from administrative insurance databases. Results Average work incapacity over the 6-year follow-up, including contributions from daily allowances and permanent losses from disability, was slightly but insignificantly higher under ICM than under SCM (21.6 vs. 21.3 % of pre-accident work capacity). Remaining work incapacity after 6 years of follow-up showed no difference between ICM and SCM (8.9 vs. 8.8 % of pre-accident work incapacity). Treatment costs were 43,500 Swiss Francs (CHF) in ICM compared to 39,800 in SCM (+9.4 %, p = 0.01). The number of care providers involved in ICM was 10.5 compared to 10.0 in ICM (+5.0 %, p SCM, but did increase healthcare consumption and treatment costs. It is concluded that the intensity of case management alone is not sufficient to improve rehabilitation and vocational reintegration of accident victims.

  19. 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.

  20. Development and test results of the Realtime Severe Accident Model 5 (RSAM5) based on the MAAP5 For the Kori 1 simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Hyuk; Lee, Myeong Soo [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Real Time Severe Accident Model (RSAM) in the Kori simulator employs the standard MAAP 5.01.1101 code (which is defined as MAAP 5.01) plus several statically linked libraries that interface with the simulator environment. The physical phenomena that can be envisioned inside the reactor vessel, the reactor coolant system (RCS), and the containment during severe accidents are comprehensively modeled by the MAAP5 code. The MAAP5 code has been known to be a reliable tool for understanding the sequence of events that occur during severe LWR accidents, evaluating the consequences of the failure of emergency systems, assessing the effects of operator interventions, and investigating the influence of design features of the RCS, containment, and safety systems on the accident consequences. The purpose of this paper is to describe the modeling of the Kori Unit 1 nuclear plant with the MAAP5 code and major outputs in the event of the SBO, SBO + SGTR, SBO + LBLOCA.

  1. Simulation technology for training in the management of severe accidents in nuclear power; Tecnologia de simulacion para entrenamiento en gestion de accidentes severos en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gil Moya, E.; Ruiz Martin, J. A.

    2012-07-01

    The objective of the project consists of the development of a module of severe accident based on the code Thermo-hydraulic MAAP and their integration in a Spanish CN training Simulator. Currently, stimulated the tools designed by Tecnatom aimed at training and assistance in the management of emergencies, complemented by the development of a dynamic interactive guides of severe accidents, thus constituting a set of aid for the operation.

  2. Differences in male and female injury severities in sport-utility vehicle, minivan, pickup and passenger car accidents.

    Science.gov (United States)

    Ulfarsson, Gudmundur F; Mannering, Fred L

    2004-03-01

    This research explores differences in injury severity between male and female drivers in single and two-vehicle accidents involving passenger cars, pickups, sport-utility vehicles (SUVs), and minivans. Separate multivariate multinomial logit models of injury severity are estimated for male and female drivers. The models predict the probability of four injury severity outcomes: no injury (property damage only), possible injury, evident injury, and fatal/disabling injury. The models are conditioned on driver gender and the number and type of vehicles involved in the accident. The conditional structure avoids bias caused by men and women's different reporting rates, choices of vehicle type, and their different rates of participation as drivers, which would affect a joint model of all crashes. We found variables that have opposite effects for the genders, such as striking a barrier or a guardrail, and crashing while starting a vehicle. The results suggest there are important behavioral and physiological differences between male and female drivers that must be explored further and addressed in vehicle and roadway design.

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

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

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

  4. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  5. Advanced ion exchange resins for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  6. Factors associated with non-return to work in the severely injured victims 3 years after a road accident: A prospective study.

    Science.gov (United States)

    Pélissier, C; Fort, E; Fontana, L; Charbotel, B; Hours, M

    2017-09-01

    Road accidents may impact victims' physical and/or mental health and socio-occupational life, particularly the capacity to return to work. The purpose of our study is to assess modifiable medical and socio-occupational factors of non-return to work in the severely injured 3 years after a road accident. Among1,168 road accidents casualties in the Rhône administrative Département of France followed for five years, 141 of the 222 severely injured (Maximal Abbreviated Injury Scale ≥ 3) aged more than 16 years who were in work at the time of the accident, reported whether they had returned to work in the 3 years following the accident. The subgroups of those who had (n=113) and had not returned to work (n=28) were compared for socio-occupational (gender, age, educational level, marital status, socio-occupational group) accident-related medical factors (type of road user, type of journey, responsibility in the accident, initial care) and post-accident medical factors (pain intensity, post-traumatic stress disorder, physical sequelae, quality of life) by using standardized tools. Severity of initial head, face and lower-limb injury, intense persistent pain, post-traumatic stress disorder, poor self-assessed quality of life and health status at 3 years were associated with non-return to work on univariate analysis. On multivariate analysis, severity of initial head and lower-limb injury, intense persistent pain at 3 years and post-traumatic stress disorder were significantly associated with non-return to work 3 years following severe road-accident injury. Post-traumatic stress disorder and chronic pain were essential modifiable medical determinants of non-return to work in the severely injured after a road accident: early adapted management could promote return to work in the severely injured. Improve early adapted treatment of pain and PTSD in the rehabilitation team should help the severely injured return to work following a road accident. Copyright © 2017 Elsevier

  7. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  8. Size distributions of airborne radionuclides from the fukushima nuclear accident at several places in europe.

    Science.gov (United States)

    Masson, Olivier; Ringer, Wolfgang; Malá, Helena; Rulik, Petr; Dlugosz-Lisiecka, Magdalena; Eleftheriadis, Konstantinos; Meisenberg, Olivier; De Vismes-Ott, Anne; Gensdarmes, François

    2013-10-01

    Segregation and radioactive analysis of aerosols according to their aerodynamic size were performed in France, Austria, the Czech Republic, Poland, Germany, and Greece after the arrival of contaminated air masses following the nuclear accident at the Fukushima Dai-ichi nuclear power plant in March 2011. On the whole and regardless of the location, the highest activity levels correspond either to the finest particle fraction or to the upper size class. Regarding anthropogenic radionuclides, the activity median aerodynamic diameter (AMAD) ranged between 0.25 and 0.71 μm for (137)Cs, from 0.17 to 0.69 μm for (134)Cs, and from 0.30 to 0.53 μm for (131)I, thus in the "accumulation mode" of the ambient aerosol (0.1-1 μm). AMAD obtained for the naturally occurring radionuclides (7)Be and (210)Pb ranged from 0.20 to 0.53 μm and 0.29 to 0.52 μm, respectively. Regarding spatial variations, AMADs did not show large differences from place to place compared with what was observed concerning bulk airborne levels registered on the European scale. When air masses arrived in Europe, AMADs for (131)I were about half those for cesium isotopes. Higher AMAD for cesium probably results from higher AMAD observed at the early stage of the accident in Japan. Lower AMAD for (131)I can be explained by the adsorption of gaseous iodine on particles of all sizes met during transport, especially for small particles. Additionally, weathering conditions (rain) encountered during transport and in Europe in March and April contributed to the equilibrium of the gaseous to total (131)I ratio. AMAD slightly increased with time for (131)I whereas a clear decreasing trend was observed with the AMADs for (137)Cs and (134)Cs. On average, the associated geometric standard deviation (GSD) appeared to be higher for iodine than for cesium isotopes. These statements also bear out a gaseous (131)I transfer on ambient particles of a broad size range during transport. Highest weighted activity levels were

  9. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  10. Effect of spray system on fission product distribution in containment during a severe accident in a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohammad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-08-15

    The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  11. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  12. Code Development on Aerosol Behavior under Severe Accident-Aerosol Coagulation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The behaviors of the larger aerosol particles are described usually by continuum mechanics. The smallest particles have diameters less than the mean free path of gas phase molecules and the behavior of these particles can often be described well by free molecular physics. The vast majority of aerosol particles arising in reactor accident analyses have behaviors in the very complicated regime intermediate between the continuum mechanics and free molecular limit. The package includes initial inventories, release from fuel and debris, aerosol dynamics with vapor condensation and revaporization, deposition on structure surfaces, transport through flow paths, and removal by engineered safety features. Aerosol dynamic processes and the condensation and evaporation of fission product vapors after release from fuel are considered within each MELCOR control volume. The aerosol dynamics models are based on MAEROS, a multi-section, multicomponent aerosol dynamics code, but without calculation of condensation. Aerosols can deposit directly on surfaces such as heat structures and water pools, or can agglomerate and eventually fall out once they exceed the largest size specified by the user for the aerosol size distribution. Aerosols deposited on surfaces cannot currently be resuspended.

  13. Aerial spraying to capture released radioactivity from NPP in a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Younus, Irfan; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of); Medard, Thiphaine [Ecole des Mines de Saint-Etienne, Daejeon (Korea, Republic of)

    2016-05-15

    The proposed strategy in this paper is the use of aqueous spray (water/foam) mixed with suitable chemical additives to capture, dissolve and stabilize the radioactive gases and aerosol particles released from leaked reactor containment and auxiliary building. The spray system can be approached to the leaked reactor building through the use of a truck with high rising cranes, unmanned aerial vehicles (UAVs, such as helicopters), aerostats, or by installing fixed piping structure around the containment building depending on the accident situation. Laboratory-scale experimental system was setup to examine the performance of such systems. The alkaline water (aqueous NaOH.Na{sub 2}S{sub 2}O{sub 3}) and foam-based spray material (sodium lauryl sulphate) were used to examine capture efficiency of gaseous iodine and aerosol particles. The gaseous iodine and aerosol removal efficiency of foam-based spray is higher when compared with alkaline water-based spray. 2. The nozzle producing full cone spray provides the better removal efficiency than nozzle producing hollow cone spray patterns.

  14. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Mun, Ch

    2007-03-15

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  15. ThermalGhydraulic Simulation of DEDVI Accident for Advanced Passive PWR%先进非能动核电厂DEDVI事故热工水力模拟分析

    Institute of Scientific and Technical Information of China (English)

    余健明; 曹学武

    2016-01-01

    The accident analysis model is established by the code of Relap5/Mod 3.4, which includes the Reactor Coolant System (RCS),simplified secondary system and Engineering Safety Features (ESF). A typical Small-Break LOCA(SBLOCA)accident, Double-Ended Direct Vessel Inj ection (DEDVI ), is selected to analyze the accident scenario and sensitivity analyses of entrainment models have been taken with respect to pressure,mass flow rate,liquid levels and peak cladding temperature. The results show that the break and ADS system can depressurize the RCS quickly and the coolant from CMT,ACC and IRWST can mitigate the accidental consequence of DEDVI effectively. Sensitivity analysis of entrainment models shows that homogenous flow model creates higher liquid discharge flow rate comparing to nonhomogenous flow model.%采用 Relap5/Mod3.4程序建立了先进非能动核电厂的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施.针对小破口失水事故(SBLOCA)中的直接安注管双端断裂事故(DEDVI)进行分析,并着重对 SBLOCA 现象识别和排序表(PIRT)中对其影响较大的液滴夹带进行敏感性分析.分析结果表明,对直接安注管双端断裂事故,破口和自动卸压系统(ADS)能够有效地使反应堆冷却剂系统降压,堆芯补水箱(CMT)、安注箱(ACC)和安全壳内置换料水箱(IRWST)能够迅速实现堆芯补水,确保堆芯冷却.对液滴夹带的敏感性分析表明,对于位置较高的第4级 ADS,喷放流量对液滴夹带模型比较敏感,使用均相流模型计算时,其液相流量显著高于非均相流模型.

  16. Guide update Severe Accident Management (SAMG) of CN. Almaraz post Fukushima; Actualizacion de las Guias de Gestion de Accidente Severo (GGAS) de CN. Almaraz post Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Fanegas, R.; Aguado Miquel, F.; Tanarro Onrubia, A.; Uruburu Rodriguez, A.

    2014-07-01

    The work is part of the activities carried out by CN. Almaraz in applying lessons learned from the Fukushima accident. The achievement of this objective requires a substantial change in the Guidelines Severe Accident Management (SAMG), starting with the adaptation of the Revision 2 of the Generic Guidelines (SAMG) Owners Group (PWROG, January 2013), which is the work is the fundamental part of this paper. (Author)

  17. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  18. Rising and boiling of a drop of volatile liquid in a heavier one: application to the LMFBR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Pigny, Sylvain L.; Coste, Pierre F. [DEN/DER/SSTH, CEA/Grenoble, 38054 Grenoble Cedex 9 (France)

    2005-07-01

    Full text of publication follows: The rising and, simultaneously the boiling, of a droplet of volatile liquid in a heavier one is computation-ally investigated. Our calculations are performed with the help of the SIMMER code, in which a specific DNS algorithm is developed, to represent surface tension between the different media in an explicit way. This is required to represent the physical contact that occurs between two liquids and the vapor from the lighter one, since interfacial heat transfers, and therefore boiling kinetics, merely depend on it. The behavior of the three fluids system is of interest as a key phenomenon related to the transition phase of LMFBR severe accidents, before the formation of a fully developed bubble column. The driven force due to the boiling of steel drops can play a major role in the relocation, and, consequently, the recriticality of UO{sub 2} fuel. The problem is investigated focusing first on analytical experiments, built-up with simulating materials, and for which accurate experimental results are provided. The dependence of results with regard to thermodynamical and physical properties is underlined. This point is of interest in view of some uncertainties in the knowledge of data concerning the materials present in the reactor at high temperature. The pressure level is a key parameter in the accident scenarios: its influence is uppermost on the volumic mass of the gas. It is also outlined. (authors)

  19. 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)

  20. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  1. 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.)

  2. 福岛第一核电厂严重事故管理研究%Research on severe accident management in Fukushima Daiichi Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    刘凯; 王炜

    2013-01-01

    The accident of Fukushima Nuclear Power Plant led to a severe accident of core meltdown, and its process of emergency management exposed various defects which raised great concern about severe accident management in nuclear power plants. In this paper, the specifications of severe accident management that issued by IAEA and Japan were overviewed. Based on Japan specifications, the analysis of sequences and management strategies were presented on severe accident in Fukushima Daiichi Nuclear Power Plant. Following identification of defects on severe accident management, possible corrective measures for current and future plants were discussed. Finally , an approach and a frame model for severe accident management were presented, which may improve nuclear safety in current and future plants.%日本福岛核事故造成了堆芯熔毁的严重事故,应急处置过程暴露出严重事故管理的种种不足,引起对核电厂严重事故管理的关注.简述了国际原子能机构和日本关于核电厂严重事故管理的规范要求,分析了福岛第一核电厂事故序列和严重事故管理策略,讨论了严重事故管理存在的问题及其可能的改进措施,最后提出了改进核电厂严重事故管理的框架模型和方法.

  3. Fukushima Daiichi accident as a stress test for the national system for the protection of the public in event of severe accident at NPP

    Directory of Open Access Journals (Sweden)

    V.A. Kutkov

    2017-03-01

    Full Text Available It is proposed that the circumstances of the Fukushima Daiichi nuclear accident on 11 March 2011 in Japan should be used as the framework for the stress test of the national system for the protection of public in the beyond design extension conditions at NPP. Stress tests of the public protection strategy show to what extent the national system is stable under the most unfavorable NPP conditions and give an understanding of the potential vulnerabilities and the ways to resolve them. A definition of the Fukushima stress test model has been provided, and the actions undertaken by Japanese authorities under the conditions of the Fukushima Daiichi accident have been considered as the response to this stress test. The stress test has revealed major vulnerabilities in the strategy for the protection of public in the event of an accident at an NPP, which was successfully proven many times by over a hundred exercises at different levels. The stress test showed that the principal vulnerability of protection strategy being in use in Japan in 2011 was the reliance on computer systems in the assessment of the emergency exposure for decision-making during the emergency response phase. It is proposed, that the Fukushima stress test should be used to identify the vulnerabilities in the Russian Federation's strategy for the protection of public in the event of a nuclear accident and to use the lessons learnt from the test results to perfect this strategy.

  4. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  5. Methodology to estimate the cost of the severe accidents risk / maximum benefit; Metodologia para estimar el costo del riesgo de accidentes severos / beneficio maximo

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza, G.; Flores, R. M.; Vega, E., E-mail: gozalo.mendoza@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    For programs and activities to manage aging effects, any changes to plant operations, inspections, maintenance activities, systems and administrative control procedures during the renewal period should be characterized, designed to manage the effects of aging as required by 10 Cfr Part 54 that could impact the environment. Environmental impacts significantly different from those described in the final environmental statement for the current operating license should be described in detail. When complying with the requirements of a license renewal application, the Severe Accident Mitigation Alternatives (SAMA) analysis is contained in a supplement to the environmental report of the plant that meets the requirements of 10 Cfr Part 51. In this paper, the methodology for estimating the cost of severe accidents risk is established and discussed, which is then used to identify and select the alternatives for severe accident mitigation, which are analyzed to estimate the maximum benefit that an alternative could achieve if this eliminate all risk. Using the regulatory analysis techniques of the US Nuclear Regulatory Commission (NRC) estimates the cost of severe accidents risk. The ultimate goal of implementing the methodology is to identify candidates for SAMA that have the potential to reduce the severe accidents risk and determine if the implementation of each candidate is cost-effective. (Author)

  6. Recent numerical simulations and experiments on coolability of debris beds during severe accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J., E-mail: joerg.starflinger@ike.uni-stuttgart.de; Buck, M.; Hartmann, A.; Kulenovic, R.; Leininger, S.; Rahman, S.; Rashid, M.

    2015-12-01

    Highlights: • Investigation on coolability of three-dimensional debris beds has been performed. • Computer code MEWA (Melt Water) is introduced and described briefly. • Validation experiments have been carried out in DEBRIS facility. • Comparison of MEWA simulations and DEBRIS experiments show good agreement. • Example simulation on reactor scale was performed to explain the analysis method. - Abstract: In the course of a severe accident in light water reactors with core degradation, so-called debris beds can be formed inside the reactor pressure vessel or in the reactor cavity. The strategy to analyse the coolability of such debris beds with both experiments and numerical simulations is discussed. The numerical simulations are carried out with MEWA (MElt WAter) code, being developed at the institute for the prediction of the thermal-hydraulic conditions inside a debris bed, including the prediction of dryout heat flux. The simulations show good agreement with experimental data of the DEBRIS experiments.

  7. In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kymaelaeinen, O.; Tuomisto, H. [IVO International Ltd., Vantaa (Finland); Theofanous, T.G. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This paper summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.

  8. Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club; Pourgol-Mohammad, Mohammad [Sahand Univ. of Technology, Tabriz (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-03-15

    This paper deals with simulation, sensitivity and uncertainty analysis of LP-FP-2 experiment of LOFT test facility. The test facility simulates the major components and system response of a pressurized water reactor during a LOCA. MELCOR code is used for predicting the fission product release from the core fuel elements in LOFT LP-FP-2 experiment. Moreover, sensitivity and uncertainty analysis is performed for different CORSOR models simulating release of fission products in severe accident calculations for nuclear power plants. The calculated values for the fission product release are compared under different modeling options to the experimental data available from the experiment. In conclusion, the performance of 8 CORSOR modeling options is assessed for available modeling alternatives in the code structure.

  9. 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)

  10. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  11. Coupled simulation of steam line break accident; Simulation couplee d'un accident de rupture de tuyauterie vapeur

    Energy Technology Data Exchange (ETDEWEB)

    Royer, E.; Raimond, E.; Caruge, D

    2000-07-01

    The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W. [and others

    1997-04-01

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

  13. The impact on the competence on severe accidents following the Fukushima event

    Energy Technology Data Exchange (ETDEWEB)

    Band, Sebastian; Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany); Schaffrath, Andreas; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2013-09-15

    Fukushima related questions are currently being addressed at Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH within several research projects funded by the Federal Ministry for Environment, Nature Conservation and Nuclear Safety (BMU) and Federal Ministry of Economics (BMWi). In the following section first results of selected issues are presented. (orig.)

  14. Do seat belts and air bags reduce mortality and injury severity after car accidents?

    Science.gov (United States)

    Cummins, Justin S; Koval, Kenneth J; Cantu, Robert V; Spratt, Kevin F

    2011-03-01

    We studied National Trauma Data Bank data to determine the effectiveness of car safety devices in reducing mortality and injury severity in 184,992 patients between 1988 and 2004. Safety device variables were seat belt used plus air bag deployed; only seat belt used; only air bag deployed; and, as explicitly coded, no device used. Overall mortality was 4.17%. Compared with the no-device group, the seat-belt-plus-air-bag group had a 67% reduction in mortality (adjusted odds ratio [AOR], 0.33; 99% confidence interval [CI], 0.28-0.39), the seatbelt- only group had a 51% mortality reduction (AOR, 0.49; 99% CI, 0.45-0.52), and the air-bag-only group had a 32% mortality reduction (AOR, 0.68, 99% CI, 0.57-0.80). Injury Severity Scores showed a similar pattern.

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

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A. [and others

    1997-04-01

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

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

    Science.gov (United States)

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

    2014-04-01

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

  18. Dysphagia and cerebrovascular accident: relationship between severity degree and level of neurological impairment.

    Science.gov (United States)

    Itaquy, Roberta Baldino; Favero, Samara Regina; Ribeiro, Marlise de Castro; Barea, Liselotte Menke; Almeida, Sheila Tamanini de; Mancopes, Renata

    2011-12-01

    The aim of this case study was to verify the occurrence of dysphagia in acute ischemic stroke within 48 hours after the onset of the first symptoms, in order to establish a possible relationship between the level of neurologic impairment and the severity degree of dysphagia. After emergency hospital admission, three patients underwent neurological clinical evaluation (general physical examination, neurological examination, and application of the National Institute of Health Stroke Scale - NIHSS), and clinical assessment of swallowing using the Protocolo Fonoaudiológico de Avaliação do Risco para Disfagia (PARD--Speech-Language Pathology Protocol for Risk Evaluation for Dysphagia). One of the patients presented functional swallowing (NIHSS score 11), while the other two had mild and moderate oropharyngeal dysphagia (NIHSS scores 15 and 19, respectively). The service flow and the delay on the patients' search for medical care determined the small sample. The findings corroborate literature data regarding the severity of the neurological condition and the manifestation of dysphagia.

  19. Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

    Energy Technology Data Exchange (ETDEWEB)

    Beuzet, Emilie, E-mail: emilie.beuzet@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Lamy, Jean-Sylvestre, E-mail: jean-sylvestre.lamy@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Bretault, Armelle, E-mail: armelle.bretault@edf.f [EDF R and D, 1 Avenue du General de Gaulle, F-92140 Clamart (France); Simoni, Eric, E-mail: simoni@ipno.in2p3.f [Institut de Physique Nucleaire, Universite Paris Sud XI, F-91406 Orsay (France)

    2011-04-15

    In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues. A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process. A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed

  20. Cytogenetical dose estimation for 3 severely exposed patients in the JCO criticality accident in Tokai-mura.

    Science.gov (United States)

    Hayata, I; Kanda, R; Minamihisamatsu, M; Furukawa, M; Sasaki, M S

    2001-09-01

    A dose estimation by chromosome analysis was performed on the 3 severely exposed patients in the Tokai-mura criticality accident. Drastically reduced lymphocyte counts suggested that the whole-body dose of radiation which they had been exposed to was unprecedentedly high. Because the number of lymphocytes in the white blood cells in two patients was very low, we could not culture and harvest cells by the conventional method. To collect the number of lymphocytes necessary for chromosome preparation, we processed blood samples by a modified method, called the high-yield chromosome preparation method. With this technique, we could culture and harvest cells, and then make air-dried chromosome slides. We applied a new dose-estimation method involving an artificially induced prematurely condensed ring chromosome, the PCC-ring method, to estimate an unusually high dose with a short time. The estimated doses by the PCC-ring method were in fairly good accordance with those by the conventional dicentric and ring chromosome (Dic+R) method. The biologically estimated dose was comparable with that estimated by a physical method. As far as we know, the estimated dose of the most severely exposed patient in the present study is the highest recorded among that chromosome analyses have been able to estimate in humans.

  1. Neutronics aspects associated to the prevention and mitigation of severe accidents in sodium cooled reactor cores; Aspects de neutronique associes a la prevention et a la reduction des accidents graves dans les coeurs de reacteurs a caloporteur sodium

    Energy Technology Data Exchange (ETDEWEB)

    Poumerouly, S.

    2010-12-15

    Among all the types of accidents to be considered for the safety licensing of a plant, some have a very low probability of occurrence but might have very important consequences: the severe accidents or Hypothetical Core Disruptive Accidents (HCDA). The studies on the scenario of these accidents are performed in parallel to the prevention studies. In this PhD report, two representative safety cases are studied: the Unprotected Loss Of Flow (ULOF) and the Total Instantaneous Blockage (TIB). The objectives are to understand what causes the reactivity increase during these accidents and to find means to reduce the energetic release of the scenario (ULOF) or to find ways to trigger the core prior to the propagation of the accident (TIB). At first, the accidents are studied in static calculations with the ERANOS code system. The accidents are divided into several steps and the reactivity insertions at each step are explained. This study shows the importance of the removal of the structures as well as of the radial leakage changes during the core slumping-down. The study also gives the amounts of fuel to be ejected or of absorber to be injected in both accidents. These values give tracks to the following more accurate studies, the transient studies. The transient studies were performed with the SIMMER code system, coupling thermo-hydraulics and neutronics. SIMMER data and algorithms have been improved so as to better predict ERANOS results (former discrepancies were up to 1.5$). The SIMMER reactivity calculation is improved by 0.8$ with variations of reactivity due to the motion of materials correctly predicted. A new algorithm for the {beta}-effective was implemented in SIMMER so as to be more accurate and easier to manage. SIMMER is then used to calculate the secondary phase of the ULOF, while the primary phase is calculated with ERANOS thanks to some assumptions. The assumptions are very much based on the fact that the movement of materials stops whenever the energy

  2. COTELS project (1): overview of project to study FCI and MCCI during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Hideo; Kato, Masami; Sakaki, Isao [Nuclear Power Engineering Corp., Tokyo (Japan). System Safety Dept.; Cherepnin, Y.; Vasilyev, Y.; Kolodeshnikov, A.; Zhdanov, V.; Zuev, V. [National Nuclear Center, Kurchatov (Kazakhstan)

    2000-05-01

    Fuel coolant interaction (FCI) and molten core concrete interaction (MCCI) have been studied experimentally within the framework of COTELS project from 1995 as a joint study between NUPEC (Japan) and NNC (Republic of Kazakhstan) using one of the testing complex at NNC. The testing complex includes three experimental facilities ''SLAVA'', ''LAVA'' and ''LAVA-M'' for debris coolability tests. Three types of experiments were carried out. To get the molten corium, the electric induction melting furnace (EMF) was used. The EMF produced {proportional_to}60 kg of corium containing UO{sub 2}, stainless steel, Zr and ZrO{sub 2}. The temperature of the produced melt was about 3200 K. The melt was discharged into the water pool in test A or onto the concrete trap in test B/C. The corium in the concrete trap was heated in test B/C by another induction melt heater. Prior to main test A and test B/C, several supporting experiments were conducted. Integrity of graphite crucible with TaC sheet during producing UO{sub 2} corium was confirmed experimentally. The induction melt heater was calibrated and the efficiency for the induction heater of ''LAVA-M'' facility was determined as 47%. The thermal conductivity and thermal diffusivity of concrete up to about 1073 K, and melting-solidification points of eutectics generated from corium components were determined experimentally. Discharge corium behavior, using UO{sub 2} corium, was also observed by speed cameras in test 01. (orig.)

  3. Calculation of Spent Fuel Pool Severe Accident With MELCOR%MELCOR 乏燃料水池严重事故计算分析

    Institute of Scientific and Technical Information of China (English)

    邓坚; 向清安; 周克峰

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MEL‐COR code to study the severe accident phenomena caused by the long term station black‐out (SBO) ,including spent fuel heatup ,zirconium cladding oxidation ,and the injection into SFP to mitigate the severe accident . The results show that the severe accident progression is slow and relates directly with the initial water level in SFP . It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident .%针对长时间全厂断电(SBO)事故,采用MELCOR程序建立了乏燃料水池的计算分析模型,研究了乏燃料组件加热升温、锆包壳氧化等严重事故现象,并计算了向乏燃料水池注水缓解严重事故的效果。研究表明:乏燃料水池内的严重事故进程相对缓慢,且与乏燃料水池初始水位直接相关;向乏燃料水池注水是缓解乏燃料水池严重事故的有效手段之一。

  4. 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.

  5. 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

  6. A qualitative study analyzing access to physical rehabilitation for traffic accident victims with severe disability in Brazil.

    Science.gov (United States)

    Sousa, Kelienny de Meneses; Oliveira, Wagner Ivan Fonsêca de; Melo, Laiza Oliveira Mendes de; Alves, Emanuel Augusto; Piuvezam, Grasiela; Gama, Zenewton André da Silva

    2017-03-01

    Purpose To identify access barriers to physical rehabilitation for traffic accident (TA) victims with severe disability and build a theoretical model to provide guidance towards the improvement of these services. Methods Qualitative research carried out in the city of Natal (Northeast Brazil), with semi-structured interviews with 120 subjects (19 key informer health professionals and 101 TA victims) identified in a database made available by the emergency hospital. The interviews were analyzed using Alceste software, version 4.9. Results The main barriers present in the interviews were: (1) related to services: bureaucratic administrative practises, low offer of rehabilitation services, insufficient information on rehabilitation, lack of guidelines that integrate hospital and ambulatory care and (2) related to patients: financial difficulties, functional limitations, geographic distance, little information on health, association with low education levels and disbelief in the system and in rehabilitation. Conclusion The numerous access barriers were presented in a theoretical model with causes related to organizational structure, processes of care, professionals and patients. This model must be tested by health policy-makers and managers to improve the quality of physical rehabilitation and avoid unnecessary prolongation of the suffering and disability experienced by TA survivors. Implications for rehabilitation Traffic accidents (TAs) are a global health dilemma that demands integrality of preventive actions, pre-hospital and hospital care and physical rehabilitation (PR). This study lays the foundation for improving access to PR for TA survivors, an issue of quality of care that results in preventable disabilities. The words of the patients interviewed reveal the suffering of victims, which is often invisible to society and given low priority by health policies that relegate PR to a second plan ahead of prevention and urgent care. A theoretical model of the

  7. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2008-01-15

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  8. A study on the overall economic risks of a hypothetical severe accident in nuclear power plant using the delphi method

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Han Ki; Kim, Joo Yeon; Lee, Jai Ki [Hanyang University, Seoul (Korea, Republic of)

    2008-12-15

    Potential economic impact of a hypothetical severe accident at a nuclear power plant(Uljin units 3/4) was estimated by applying the delphi method, which is based on the expert judgements and opinions, in the process of quantifying uncertain factor. For the purpose of this study, it is assumed that the radioactive plume directs the inland direction. Since the economic risk can be divided into direct costs and indirect effects and more uncertainties are involved in the latter, the direct costs were estimated first and the indirect effects were then estimated by applying a weighting factor to the direct cost. The delphi method however subjects to risk of distortion or discrimination of variables because of the human behavior pattern. A mathematical approach based on the Bayesian inferences was employed for data processing to improve the delphi results. For this task, a model for data processing was developed. One-dimensional Monte Carlo analysis was applied to get a distribution of values of the weighting factor. The mean and median values of the weighting factor for the indirect effects appeared to be 2.59 and 2.08, respectively. These values are higher than the value suggested by OECD/NEA, 1.25. Some factors such as small territory and public attitude sensitive to radiation could affect the judgement of panel. Then the parameters of the model for estimating the direct costs were classified as U- and V-types, and two-dimensional Monte Carlo analysis was applied to quantify the overall economic risk. The resulting median of the overall economic risk was about 3.9% of the Gross Domestic Products (GDP) of Korea in 2006. When the cost of electricity loss, the highest direct cost, was not taken into account, the overall economic risk was reduced to 2.2% of GDP. This assessment can be used as a reference for justifying the radiological emergency planning and preparedness.

  9. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents: Year 2004 Activity. APRI 5 report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Park, H.S.; Nayak, A.K.; Hansson, R.C.; Chiferaw, D.; Stepanyan, A.; Rao, R.S.; Karbojian, A. [Royal Inst. of Technology, Stockholm (Sweden). Div. of Nuclear Power Safety

    2005-04-01

    This report presents descriptions of the major results obtained in the research program 'Melt-Structure-Water Interaction (MSWI)' at NPS/RIT during the year 2004. The primary objectives of the MSWI Project in year 2004 were to study (1) the in-vessel and exvessel melt/debris bed coolability process when melt is flooded with water, and (2) the energetics and characteristics of steam explosions. Our general approaches are to establish scaling relationships so that the data obtained in the experiments could be extended to prototypical accident geometries and conditions, develop phenomenological or computational models for the processes under investigation and validate the existing and newly developed models against data obtained at RIT and at other laboratories. In 2004, several experimental programs, such as the COMECO (Corium MElt COolability), POMECO (POrous MEdia COolability) and MISTEE (Micro-Interactions in STeam Explosion Experiments) programs were continued. The SIMECO (Simulation of MElt Coolability) program was restarted in 2004. The construction of the POMECO-GRAND (POrous MEdia COolability) facility was delayed due to lack of finances. However, existing POMECO facility was modified to study 3-D effects on debris coolability. In this report, the results from the COMECO experiment with high temperature oxidic melt, from the POMECO experiments for the multi-dimensional effects on debris bed coolability, from the SIMECO experiment for three-layer pool configuration and from the MISTEE experiments for steam explosion characteristics and loads are described. For analytical efforts, results from the COMETA code for the entire process of the steam explosions are discussed.

  10. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    Energy Technology Data Exchange (ETDEWEB)

    Hermsmeyer, S. [European Commission JRC, Petten (Netherlands). Inst. for Energy and Transport; Herranz, L.E.; Iglesias, R. [CIEMAT, Madrid (Spain); and others

    2015-07-15

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  11. 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.

  12. Suppression Pools: paradigm of the thermalhydraulic effect on severe accidents; Piscinas de Supresion: Paradigma del efecto de la thermohidraulica durante accidentes severos

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez del Pra, C.

    2016-08-01

    Influence of thermal-hydrualic phenomena on severe accident unforlding is beyond question. The present paper supports this statement on two key aspects of a severe accident: preservation of containment integrity and transport of fission products once released from fuel. To illustrate them, the attention is focused on suppression pools performance and, particularly, on some recent findings stemming from authors research of Fukushima scenarios. Gas behvaior at the injection point and its later evolution, potential axial and/or azimuthal stratification of the aqueous body or water saturation state, are some of the processes tha more strongly affect the role of pools as a mass and energy sink. They are described and discussed in detail. (Author)

  13. Vessel-related problems in severe accidents, International Research Projects; La problematica de la vasija en los accidentes severos. Proyectos internacionales de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Figueras, J. M. [Consejo de Seguridad Nuclear. Madrid (Spain)

    2000-07-01

    The paper describes those most relevant aspects of research programmes and projects, on the behavior of vessel during severe accidents with partial or total reactor core fusion, performed during the last twenty years or still on-going projects, by countries or international organizations in the nuclear community, presenting the most important technical aspects, in particular the results achieved, as well as the financial and organisational aspects. The paper concludes that, throughout a joint effort of the international nuclear community, in which Spain has been present via private and public organizations, actually exist a reasonable technical and experimental knowledge of the vessel in case of severe accidents, but still there are aspects not fully solved which are the basis for continuing some programmes and for proposal of new ones. (Author)

  14. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  15. PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-10-01

    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

  16. Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment

    Directory of Open Access Journals (Sweden)

    Sang-Won Lee

    2014-01-01

    Full Text Available After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to 10-3 considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.

  17. Material effects on multiphase phenomena in late phases of severe accidents of nuclear reactors; Effets des materiaux sur les phenomenes multiphasiques se produisant lors des phases avancees d'accident grave de reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M.; Froment, K

    2003-07-01

    This paper reviews and presents work carried out in the French Atomic Energy Commission (CEA) on the subject of nuclear severe accidents, i.e. those which are accompanied by melting of the nuclear core material. The emphasis is on the (crucial) thermodynamic and material behaviour of corium melts in the solidus-liquidus temperature interval, which is linked to the thermal hydraulic description. A global model approach is proposed. The work is presented in the context of the overall international effort in the area. (authors)

  18. Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, V.; Collins, B.L.; Sackett, K.E.; Coppin, C.E.

    1978-02-01

    Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of these tests was to refine the definition of the upper limit of steam generator tube ruptures at which high peak cladding temperatures occur, as set by Test S-28-1. During these tests, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant in a PWR. Thirty (Test S-28-7), 34 (Test S-28-9), and 20 (Test S-28-12) steam generator tube ruptures were simulated by a controlled injection from a heated accmulator into the intact loop hot leg.

  19. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  20. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

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

    Science.gov (United States)

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

    2014-06-01

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

  2. Applying Functional Modeling for Accident Management of Nucler Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    The paper investigates applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented....

  3. Multidisciplinary treatment for a young patient with severe maxillofacial trauma from a snowmobile accident: a case report.

    Science.gov (United States)

    Yamano, Seiichi; Nissenbaum, Mark; Dodson, Thomas B; Gallucci, German O; Sukotjo, Cortino

    2010-01-01

    Abstract This clinical report describes the oral rehabilitation of a 15-year-old male patient who was involved in a snowmobile accident and suffered multiple mid-face and mandibular fractures. Consequences of the accident included avulsion of teeth numbers 5 to 10 and 21 to 26, and a significant amount of maxillary and mandibular anterior alveolar bone loss. The patient underwent open reduction and rigid fixation of the fractured left zygoma, comminuted LeFort I maxillary fracture, and left body of the mandible; closed reduction of the bilateral condylar fractures; autologous corticocancellous bone grafting to the maxilla and mandible; implant placement; and prosthesis fabrication. This multidisciplinary approach successfully restored function and esthetics.

  4. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  5. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  6. Performance and scenario evaluation of PAFS through the LOFW accident in APR1400 by using MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Bae, Byoung Uhn; Yun, Byong Jo [Korea Atomic Energy Institute, Daejeon (Korea, Republic of)

    2009-07-01

    In order to enhance the safety feature of the APR1400 through the passive ways, the passive auxiliary feedwater system(PAFS) is under preliminary consideration by KAERI. For the successful adaptation of PAFS, accident scenario evaluation of PWR plant that is assumed to have the PAFS system should be performed. Condensing heat exchanger assemblies are installed at the exterior boundary of the containment building per one steam generator. The performance of the heat exchanger is designed to remove the decay heat of the fuel completely. In normal operation condition, PAFS system is not connected with the steam and feed lines. A Total Loss of Feed Water(TLOFW) accident is selected for the performance and scenario evaluation after the severity check. The PAFS connection valves are open at the signal of 25% level trip of steam generator. With the single failure assumption of PAFS open valve, the scenario propagations are calculated by using MARS code.

  7. Research in the Ciemat on severe accidents: strategy and recent results; Investigaciones en el Ciemat sobre accidentes severos: estrategia y resultados recientes

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.

    2012-11-01

    Severe accident research is a fundamental brick in the nuclear technology wall. Its complexity entails huge challenges that require international cooperation to be overcome. CIEMAT has accumulated more than 40 years of experience in the field. By setting a structured research strategy and a continuous enhancement of theoretical an experimental capabilities, CIEMAT has recently produced the results on which this article builds up. Through them, both its working domains and its firm commitment for a continuous growth of knowledge and know-how are outlined. (Author) 24 refs.

  8. Decreasing adhesions and avoiding further surgery in a pediatric patient involved in a severe pedestrian versus motor vehicle accident

    Directory of Open Access Journals (Sweden)

    Amanda D. Rice

    2014-02-01

    Full Text Available In this case study, we report the use of manual physical therapy in a pediatric patient experiencing complications from a life-threatening motor vehicle accident that necessitated 19 surgeries over the course of 12 months. Post-surgical adhesions decreased the patient’s quality of life. He developed multiple medical conditions including recurrent partial bowel obstructions and an ascending testicle. In an effort to avoid further surgery for bowel obstruction and the ascending testicle, the patient was effectively treated with a manual physical therapy regimen focused on decreasing adhesions. The therapy allowed return to an improved quality of life, significant decrease in subjective reports of pain and dysfunction, and apparent decreases in adhesive processes without further surgery, which are important goals for all patients, but especially for pediatric patients.

  9. Evaluation of potential severe accidents during low power and shutdown operations at Surry: Unit 1, Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Pratt, W.T. [eds.; Musicki, Z. [Brookhaven National Lab., Upton, NY (United States)

    1995-10-01

    This document contains a summarization of the results and insights from the Level 1 accident sequence analyses of internally initiated events, internally initiated fire and flood events, seismically initiated events, and the Level 2/3 risk analysis of internally initiated events (excluding fire and flood) for Surry, Unit 1. The analysis was confined to mid-loop operation, which can occur during three plant operational states (identified as POSs R6 and R10 during a refueling outage, and POS D6 during drained maintenance). The report summarizes the Level 1 information contained in Volumes 2--5 and the Level 2/3 information contained in Volume 6 of NUREG/CR-6144.

  10. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  11. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  12. Who by accident? The social morphology of car accidents.

    Science.gov (United States)

    Factor, Roni; Yair, Gad; Mahalel, David

    2010-09-01

    Prior studies in the sociology of accidents have shown that different social groups have different rates of accident involvement. This study extends those studies by implementing Bourdieu's relational perspective of social space to systematically explore the homology between drivers' social characteristics and their involvement in specific types of motor vehicle accident. Using a large database that merges official Israeli road-accident records with socioeconomic data from two censuses, this research maps the social order of road accidents through multiple correspondence analysis. Extending prior studies, the results show that different social groups indeed tend to be involved in motor vehicle accidents of different types and severity. For example, we find that drivers from low socioeconomic backgrounds are overinvolved in severe accidents with fatal outcomes. The new findings reported here shed light on the social regularity of road accidents and expose new facets in the social organization of death. © 2010 Society for Risk Analysis.

  13. Design Study of Nuclear Power Plant Severe Accidents Monitoring and Control System%核电厂严重事故监测和控制系统的设计研究

    Institute of Scientific and Technical Information of China (English)

    杜德君; 何庆镭

    2015-01-01

    HAF102-2004 "Nuclear power plant safety requirements for quality assurance" requires: besides of design reference, nuclear power plant design must consider the specific case beyond design reference, which includes the behavior of selected severe accidents. After the Fukushima accidents, each country pays more attention to sever accidents, the prevention and mitigation of severe accidents becomes one important point for the nuclear plant design. In the third generation nuclear power plant, the severe accidents monitoring and control system should be implemented to realize the function of prevention and mitigation for severe accidents.%HAF102-2004《核动力厂设计安全规定》中要求:除了设计基准外,设计中还必须考虑核动力厂在特定的超设计基准事故包括选定的严重事故中的行为。2011年福岛事故后,各国对严重事故更加关注,严重事故的预防和缓解成为核电厂设计中的一个重点。在三代核电厂设计中,增加了专门的严重事故监测和控制系统用来实现严重事故的预防和缓解功能。

  14. 重大工艺爆炸事故严重度评价%SEVERITY EVALUATION OF MAJOR PROCESS EXPLOSION ACCIDENT

    Institute of Scientific and Technical Information of China (English)

    王三明; 蒋军成

    2001-01-01

    Severity evaluation and models of major explosion process accident have been put forward on the basis of study on many typical hazards evaluation models. With the models the software called DANGER for severity evaluation of major process explosion accident has been developed. The design of function modules of the software has been introduced. Two evaluation cases of vapor cloud explosion and boiling liquid expanding vapor cloud explosion have been provided.%在研究分析了许多典型国内外事故危险性评价模型的基础上,总结并提出了重大工艺爆炸事故危险性分级、严重度评价方法及模型。并利用此模型开发了重大工艺爆炸事故严重度评价软件,介绍了评价软件的功能模块设计。并列举了评价软件对重大蒸气云爆炸、液化气和过热液体扩展蒸气爆炸事故的评价实例。

  15. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  16. 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)

  17. Severe Psychological Distress of Evacuees in Evacuation Zone Caused by the Fukushima Daiichi Nuclear Power Plant Accident: The Fukushima Health Management Survey

    Science.gov (United States)

    Kunii, Yasuto; Suzuki, Yuriko; Shiga, Tetsuya; Yabe, Hirooki; Yasumura, Seiji; Maeda, Masaharu; Niwa, Shin-ichi; Otsuru, Akira; Mashiko, Hirobumi; Abe, Masafumi

    2016-01-01

    Background Following the Great East Japan Earthquake on March 11, 2011, the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant has continued to affect the mental health status of residents in the evacuation zone. To examine the mental health status of evacuee after the nuclear accident, we conducted the Mental Health and Lifestyle Survey as part of the ongoing Fukushima Health Management Survey. Methods We measured mental health status using the Kessler 6-item psychological distress scale (K6) in a total of 73,569 (response rate: 40.7%) evacuees aged 15 and over who lived in the evacuation zone in Fukushima Prefecture. We then dichotomized responders using a 12/13 cutoff on the K6, and compared the proportion of K6 scores ≥13 and ≤12 in each risk factor including demographic information, socioeconomic variables, and disaster-related variables. We also performed bivariate analyses between mental health status and possible risk factors using the chi-square test. Furthermore, we performed multivariate regression analysis using modified Poisson regression models. Results The median K6 score was 5 (interquartile range: 1–10). The number of psychological distress was 8,717 (14.6%). We found that significant differences in the prevalence of psychological distress by almost all survey items, including disaster-related risk factors, most of which were also associated with increased Prevalence ratios (PRs). Additionally, we found that psychological distress in each evacuation zone was significantly positively associated with the radiation levels in their environment (r = 0.768, p = 0.002). Conclusion The earthquake, tsunami and subsequent nuclear accident likely caused severe psychological distress among residents in the evacuation zone in Fukushima Prefecture. The close association between psychological distress and the radiation levels shows that the nuclear accident seriously influenced the mental health of the residents, which might be exacerbated by

  18. Severe Psychological Distress of Evacuees in Evacuation Zone Caused by the Fukushima Daiichi Nuclear Power Plant Accident: The Fukushima Health Management Survey.

    Directory of Open Access Journals (Sweden)

    Yasuto Kunii

    Full Text Available Following the Great East Japan Earthquake on March 11, 2011, the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant has continued to affect the mental health status of residents in the evacuation zone. To examine the mental health status of evacuee after the nuclear accident, we conducted the Mental Health and Lifestyle Survey as part of the ongoing Fukushima Health Management Survey.We measured mental health status using the Kessler 6-item psychological distress scale (K6 in a total of 73,569 (response rate: 40.7% evacuees aged 15 and over who lived in the evacuation zone in Fukushima Prefecture. We then dichotomized responders using a 12/13 cutoff on the K6, and compared the proportion of K6 scores ≥13 and ≤12 in each risk factor including demographic information, socioeconomic variables, and disaster-related variables. We also performed bivariate analyses between mental health status and possible risk factors using the chi-square test. Furthermore, we performed multivariate regression analysis using modified Poisson regression models.The median K6 score was 5 (interquartile range: 1-10. The number of psychological distress was 8,717 (14.6%. We found that significant differences in the prevalence of psychological distress by almost all survey items, including disaster-related risk factors, most of which were also associated with increased Prevalence ratios (PRs. Additionally, we found that psychological distress in each evacuation zone was significantly positively associated with the radiation levels in their environment (r = 0.768, p = 0.002.The earthquake, tsunami and subsequent nuclear accident likely caused severe psychological distress among residents in the evacuation zone in Fukushima Prefecture. The close association between psychological distress and the radiation levels shows that the nuclear accident seriously influenced the mental health of the residents, which might be exacerbated by increased risk perception. To

  19. 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.

  20. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

  1. Neutron noise measurements on Bugey 2 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Marini, J.; Romy, D.; Spadi, J.C.; Assedo, R.; Castello, G.

    1982-01-01

    Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurements runs during the first cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

  2. Smart Sensing of the Aux. Feed-water Pump Performance in NPP Severe Accidents Using Advanced GMDH Method

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In order to develop and verify the models, a number of data obtained by simulating station black out (SBO) scenario for the optimized power reactor 1000 (OPR1000) using MARS code were used. Most of monitoring systems for component have been suggested by using the directly measured data. However, it is very difficult to acquire data related to safety-critical component' status. Therefore, it is necessary to develop the new method that combines the data-based equipped with learning system and data miming technique. Many data-based modeling methods have been applied successfully to nuclear engineering area, such as signal validation, plant diagnostics and event identification. Also, the data miming is the process of analyzing data from different perspectives and summarizing it into useful information. In this study, the smart sensing technique was developed using advanced group method of data handing (GMDH) model. The original GMDH is an inductive self organizing algebraic model. The advanced GMDH model is equipped with a fuzzy concept. The proposed advanced GMDH model enhances the original GMDH model by reducing the effect of outliers and noise. The advanced GMDH uses different weightings according to their importance which is specified by the fuzzy membership grade. The developed model was verified using SBO accident simulation data for the OPR1000 nuclear power plant acquired with MARS code. Also, the advanced GMDH model was trained using the simulated development data and verified with simulated test data. The development and test data sets were independent. The simulation results show that the performance of the developed advanced GMDH model was very satisfactory, as shown in Table 1. Therefore, if the developed model can be optimized using diverse and specific data, it will be possible to predict the performance of Aux. feed water pump accurately.

  3. A study on the effect of containment filtered venting system to off-site under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Ju Young; Kwon, Tae Eun; Lee, Jai Ki [Hanyang University, Seoul (Korea, Republic of)

    2015-12-15

    The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1 and 2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0⁓35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2⁓3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.

  4. Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T.; Dinh, T.N.; Bui, V.A.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

    1995-07-01

    Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

  5. Accident resistant transport container

    Science.gov (United States)

    Andersen, John A.; Cole, James K.

    1980-01-01

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

  6. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su, E-mail: camila@lasme.coppe.ufrj.b, E-mail: gabrielromero@lasme.coppe.ufrj.b, E-mail: sujian@lasme.coppe.ufrj.b [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Nuclear Engineering Program

    2010-07-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-{omega} based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  7. Using the Star CCM+ software system for modeling the thermal state and natural convection in the melt metal layer during severe accidents in VVER reactors

    Science.gov (United States)

    Kochetov, N. A.; Loktionov, V. D.; Sidorov, A. S.

    2015-09-01

    The possibility of using the Star CCM+ software system for analyzing the thermal state of the melt pool metal layer generated as a result of melt stratification during a severe accident in pressure-vessel nuclear reactors is considered. In order to verify and substantiate the possibility of using this software system for modeling the natural convection processes in the melt at high values of the Rayleigh number, test problems were solved. The obtained results were found to be in good agreement with the known solutions and with the experimental data. The behavior of the melt metal layer was subjected to a parametric analysis for different melt heating conditions, the results of which showed that certain parameters have a determining influence on the so-called focusing effect and on the specific features of current in this layer.

  8. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.).

  9. Estimation of thermal loads on the VVER vessel under conditions of inversion of the stratified molten pool in a severe accident

    Science.gov (United States)

    Loktionov, V. D.; Mukhtarov, E. S.

    2016-09-01

    Analysis of the thermal state of molten pools that can be formed on the vessel bottom of the VVER-600 medium-power reactor during a severe anticipated accident with melting of the core is represented. Two types of the molten pool of core materials, with the two-layer and inverse three-layer stratification, are considered. Thermal loads acting on the reactor vessel from the melt are estimated depending on its formation time. Features of the thermal state of the melt in the case of its inverse stratification are analyzed. It is shown that thermal loads on the reactor vessel exceed the critical heat flux (CHF) when forming the two-layer stratified molten pool 10 and 24 h after its shutdown, and the thermal load is close to the corresponding CHF or somewhat exceeds it in 72 h. In the case of the formation of the inverse structure of the melt, one can observe a decrease by more than 2.5 times (in comparison with the two-layer stratified structure) in the thermal load on the reactor vessel in the region of its contact with the upper layer of the steel melt. Analysis of results showed that maximum densities of heat flux to the reactor vessel from the bottom metallic layer with the melt inversion did not exceed corresponding CHFs 24 and 72 h after the reactor shutdown. Because the thermal load on the reactor vessel can be localized in the region of its bottom, where the CHF is relatively small, during the inverse stratification of the melt, there is a need to carry out further in-depth experimental and analytical investigations of conditions for formation of the stratified molten pool and to obtain corrected experimental CHFs for conditions and outlines of cooling the external surface of the VVER-600 vessel in a severe accident.

  10. 严重事故条件下堆芯升温模拟%Simulation of Core Heating up During Severe Accident Sequence

    Institute of Scientific and Technical Information of China (English)

    王佳赟; 樊普

    2012-01-01

    The core heating up of API000 reactor during a severe accident sequence was simulated numerically by FLUENT. The objective was to study the uniformity of the heating up after the uncover but before significant melting of the core in more detail than that was possible using integral severe accident codes and obtain the temperature of shroud and baffle, also to assess the MAAP core heating up calculation. The results show that before significant core damaging, the shroud and baffle have melted causing an side relocation of the debris. Furthermore, the MAAP calculation of core heating up is also acceptable.%使用FLUENT计算流体程序数值模拟了AP1000在严重事故条件下的堆芯升温过程,目的是对堆芯裸露后并在其显著熔化前对堆芯升温的均匀程度进行比一体化事故程序MAAP更为详尽的研究,进行围筒和吊篮温度分析,同时评估MAAP程序堆芯升温计算结果.分析结果表明:在堆芯显著熔化时刻,堆芯围筒和吊篮已熔化,因此熔融堆芯将从侧面迁移进入下封头,同时对比证明MAAP程序关于堆芯升温的计算结果也是可接受的.

  11. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Jansen, J.M.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States); Lau, W.; Salyer, W.D. [Reliability and Performance Associates (United States)

    1993-12-01

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

  13. Organic chemistry and radiochemistry: study of chemical interactions between iodine and paint of French nuclear reactor in a severe accident situation; Chimie organique et radiochimie. Etude des interactions chimiques iode-peinture dans un reacteur nucleaire (de type R.E.P.) en situation d'accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Aujollet, Y. [Direction Generale de la Surete Nucleaire et de la Radioprotection, 75 - Paris (France)

    2005-01-01

    In Phebus (French in pile facility; PWR scale 1/5000) experiments, performed by the Institut de Radioprotection et de Surete Nucleaire, few quantities of organic iodides were registered after interaction between iodine and reactor containment paint. This study concerns all mechanisms of chemical reactions between iodine and the polymer of the paint in order to estimate the organic iodides released from the paint. At first, all the paint components had been identified. Several models of chemical sites of the polymer were synthesized and tested with iodine in different conditions of temperature and radiation. These experiments showed interactions between iodine and secondary or tertiary amines by charge transfer. In few cases, the complex of tertiary amines creates oxidation reactions. (author)

  14. 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.

  15. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  16. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew S. (Texas A& M University, College Station, TX); Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

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

  19. Development of a fission product transport module predicting the behavior of radiological materials during sever accidents in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyung Seok; Rhee, Bo Wook; Kim, Dong Ha [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ±6%. It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

  20. Seismic Shaking Table Requirements and Consideration of Fluid-Structure Interaction Effect in Seismic Response Analysis Model for In-Reactor Fuel Assembly Under Severe Earthquake Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Yoon, Kyungho; Kang, Heungsoek; Lee, Youngho; Kim, Hyungkyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Dynamic response of fuel assembly can be significantly affected by added hydrodynamic mass and additional damping from the fluid and flow inside operating reactor core. Added mass or hydrodynamic virtual mass from surrounding fluid medium can be theoretically estimated by the potential flow theory. Solving Laplace equation in terms of velocity potential can leads to calculate mass components in the mass matrix of simplified fuel FE model. Additional damping from the fluid and the flow inside reactor core are originated from fluid drag and flow lift force, respectively. Lift force from axial flow can increase fuel assembly damping by twice compared to still fluid damping from the loop testing. In practice, fuel assembly damping should be measured by mockup loop testing and referred to published data in the literature. The justification is performed via time history analysis with simplified dynamic model using a group of fuel assembly in the core. Key check points in this analysis might be the integrity of intermediate spacer grids when impacting fuels into core shroud plate or into neighboring fuel assembly. Thus, dynamic displacement and impact force at grid elevations are the important structural parameters to be traced out during the analysis and the simulation testing. KAERI have a plan to develop dynamic analysis model and to setup test infrastructure for full scale and several fuel assembly rows seismic simulation testing. This paper briefly discuss on the reference earthquake accident scenario, shaking table requirements for full-scale seismic simulation testing, virtual testing issues before the hardware setup, and modelling issue related to fluid-structure interaction effect in accident core analysis.

  1. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  2. Evaluation of passive autocatalytic recombiners (PARS) performance for a PWR-konvoi containment type with Gothic 8.1 code; Evaluacion de la implementacion de recombinadores autocataliticos pasivos (PAR) en una contencion tipo Konvoi con el codigo Gothic 8.1

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Alonso Conty, E.; Papini, D.; Jimenez Varas, G.

    2016-08-01

    The study presented in this work analyses the evaluation of Passive Autocatalytic Recombiners (PARs) performance for a PWR-Konvoi containment type as a result of an international collaboration between the Paul Scherrer institute (PSI) and the Universidad Politecnica de Madrid (UPM). The implementation study analyzes the size, location and number of the PARs to minimize the risk arising from a hydrogen release and its distribution in the containment building during a hypothetical severe accident. A detailed 3D model of containment was used for the simulations developed for the Gothic 8.1 code. In the first place, the hydrogen preferential pathways and points of hydrogen accumulation were studies and identified starting from the base case scenario without any mitigation measure. The severe accident scenario chosen is a fast release of hydrogen-steam mixture from hot leg creep rupture during SBO (Station Black-Out) accident. Secondly a configuration of PARs was simulated under the same conditions of the unmitigated case. The PAR configuration offered an improvement in the chosen accident scenario, decreasing the hydrogen concentration values below the flammability limit /hydrogen concentration below 7%) in all the containment compartments. (Author)

  3. VERIFIKASI KECELAKAAN HILANGNYA ALIRAN AIR UMPAN PADA REAKTOR DAYA PWR MAJU

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Reactor Technology and Nuclear Safety as a Technical Support Organization (TSO in terms of reactor safety verification, the verification activities have been carried out for the AP1000 that begins with failure of secondary coolant accident verification. The activity started with the technical safety features modeling such as passive core cooling system consisting of a Passive Residual Heat Removal system (PRHR, Core Makeup Tank (CMT, and In-containment Refueling Water Storage Tank (IRWST. The failure of secondary coolant accident selected is the loss of main feedwater flow to one of the steam generator simulated using the calculation program RELAP5/SCDAP/Mod3.4. The objective of analysis is to obtain sequences of changes in the thermalhydraulic parameters in the reactor due to the selected event. Analysis results obtained are validated and compared with the analysis results using the calculation program LOFTRAN in the AP1000 safety design document. The verification results show that the loss of feed-water supply has no impact on core damage, the reactor coolant system, as well as secondary systems. The ability of heat exchanger PRHR has been verified to dissipate decay heat of the core after reactor trip. Validation with the AP1000 safety design document shows compliance on most thermal hydraulic parameters. In general, the advanced PWR model equipped with passive core cooling system that has been developed remains safe in the event of loss of secondary coolant flow accident. Keywords: Verification, loss of feed water flow, AP1000

  4. Do Cognitive Models Help in Predicting the Severity of Posttraumatic Stress Disorder, Phobia, and Depression after Motor Vehicle Accidents? A Prospective Longitudinal Study

    Science.gov (United States)

    Ehring, Thomas; Ehlers, Anke; Glucksman, Edward

    2008-01-01

    The study investigated the power of theoretically derived cognitive variables to predict posttraumatic stress disorder (PTSD), travel phobia, and depression following injury in a motor vehicle accident (MVA). MVA survivors (N = 147) were assessed at the emergency department on the day of their accident and 2 weeks, 1 month, 3 months, and 6 months…

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

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

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

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

  8. Development of a three-dimensional model and calculation code for the packed bed simulation for safety analyses of severe reactor accidents; Entwicklung eines dreidimensionalen Modells und Rechencodes zur Simulation von Schuettbetten fuer Sicherheitsanalysen von schweren Reaktorstoerfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Berkhan, Ana; Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2013-07-01

    The computer code MEWA is used for the description of severe accident sequences in light-water reactors. During the reactor accident with core disruption the solidified core fragments are displaced into the lower plenum of the reactor pressure vessel (RPV) or in case of RPV failure into the water filled reactor sump. For the progress or cessation of the severe accident the cooling of the packed bed is of main importance. With the 3D version of the code it is possible to study spatially complex packed beds with respect to their coolability. Further extension of the MEWA code will include the optimization for the improvement of the calculation efficiency and reduction of computation time. The validation will be performed by re-calculation of experiments (for instance DEBRIS experiments at the IKE) and the comparison with results of the 2D version.

  9. Study of top reflooding in case of severe accident and in particular oxidation of Uranium, Zirconium, Oxygen melts; Etude du renoyage par le haut en cas d'accident grave et en particulier oxydation des melanges (U,Zr,O)

    Energy Technology Data Exchange (ETDEWEB)

    Brunet-Thibault, E

    2006-12-15

    In 1979, the Three Mile Island (TMI) accident occurred in United States and accelerated research activities in the field of severe accidents. Severe accident management procedures imply massive water injections to flood the core. The work of this thesis bent principally over this reflooding. The first part of the study concerns the core oxidation enhancement during the reflooding phase which leads to a rough increase of the concentration of burnable hydrogen in the containment. This is why the study carried on the analysis of the contribution of the oxidation of U-Zr-O mixtures, towards the total production of hydrogen during reflooding. In the second part, the study concerns top flooding modelling i.e.: with injection of water in the hot legs. Here, we attempted to define bases and realize a model allowing to describe this type of reflooding. These models were validated on the simulation of the parameter with MAAP4 code. (author)

  10. Impact of Aliquat {sup registered} 336 addition on organic iodine retention in containment-venting-scrubbing solutions for mitigation of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, A.; Zeh, P.; Buhlmann, S. [AREVA NP GmbH, Erlangen (Germany)

    2013-07-01

    To mitigate severe accident situations Filtered Containment Venting Systems have been designed, internationally qualified and implemented in modern nuclear power plants (NPPs) in order to minimize radionuclide release to environment in case of containment pressure reduction via venting. Main focus was given to the reliable and efficient aerosol retention. In addition also efficient iodine retention was requested, as this element has significant activity content in nuclear fuel in combination with high volatility and radiotoxicity. Therefore, effort is made to reduce the iodine activity in venting gases. State-of-the-art containment venting scrubbing solutions use a solution of sodium hydroxide and sodium thiosulfate in order to wash out volatile iodine species. With such a solution high retention efficiencies for elemental iodine and hydrogen iodide are achieved. Nevertheless, the retention of organic iodine species in this solution is not satisfying and the search for improvements is ongoing. A possible additive presented in literature is Aliquat {sup registered} 336 promising improved retention of volatile organic iodine species in scrubbing solutions. This Aliquat {sup registered} 336 is a water insoluble quaternary ammonium chloride salt made by the methylation of mixed tri-octyl/decyl amine. The effectiveness of such an additive was tested at elevated temperatures and pressures simulating containment venting conditions. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

  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 E (Sections E.1--E.8). Volume 2, Part 3A

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful.

  13. 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...

  14. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination