WorldWideScience

Sample records for severe accident scenarios

  1. Dose calculations for severe LWR accident scenarios

    International Nuclear Information System (INIS)

    Margulies, T.S.; Martin, J.A. Jr.

    1984-05-01

    This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well

  2. Analyses of severe accident scenarios in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.; Rimkevicius, S.; Uspuras, E.; Urbonavicius, E.

    2006-01-01

    Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena. (orig.)

  3. Shipping container response to three severe railway accident scenarios

    International Nuclear Information System (INIS)

    Mok, G.C.; Fischer, L.E.; Murty, S.S.; Witte, M.C.

    1998-01-01

    The probability of damage and the potential resulting hazards are analyzed for a representative rail shipping container for three severe rail accident scenarios. The scenarios are: (1) the rupture of closure bolts and resulting opening of closure lid due to a severe impact, (2) the puncture of container by an impacting rail-car coupler, and (3) the yielding of container due to side impact on a rigid uneven surface. The analysis results indicate that scenario 2 is a physically unreasonable event while the probabilities of a significant loss of containment in scenarios 1 and 3 are extremely small. Before assessing the potential risk for the last two scenarios, the uncertainties in predicting complex phenomena for rare, high- consequence hazards needs to be addressed using a rigorous methodology

  4. Prediction of Transient Scenarios Using AI After Severe Accident Occurrence

    International Nuclear Information System (INIS)

    Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun

    2017-01-01

    We predicted the core uncovery time, the time that core exist temperature (CET) exceeds 1200 .deg. F, reactor vessel (RV) failure time and containment failure time by using the cascaded support vector regression (SVR) model. The proposed algorithms were trained and verified using the simulation data of MAAP code for the optimized power rector (OPR1000). In this study, we predicted transient scenarios by CSVR. The MAAP code was used to describe the accident situation and the 13 measured signal data was acquired and used. The CSVR model was developed to find out the transient scenarios by using short timeintegrated signals after reactor trip. The results show that the CSVR models can predict the transient scenarios accurately.

  5. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A.

    1990-08-01

    This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

    2016-01-01

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

  8. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  9. Radionuclide release calculations for selected severe accident scenarios. PWR, ice condenser design

    Energy Technology Data Exchange (ETDEWEB)

    Denning, R S; Gieseke, J A; Cybulskis, P; Lee, K W; Jordan, H; Curtis, L A; Kelly, R F; Kogan, V; Schumacher, P M

    1986-07-01

    This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with an ice-condenser containment. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the U.S. Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. The Sequoyah Plant has been used in this study as an example of a PWR ice-condenser plant. (author)

  10. Radionuclide release calculations for selected severe accident scenarios. Volume 3. PWR, subatmospheric containment design

    International Nuclear Information System (INIS)

    Denning, R.S.; Gieseke, J.A.; Cybulskis, P.; Lee, K.W.; Jordan, H.; Curtis, L.A.; Kelly, R.F.; Kogan, V.; Schumacher, P.M.

    1986-07-01

    This report presents results of analyses of the enviromental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with a subatmospheric containment design. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. the Surry plant has been used in this study as the reference plant for a subatmospheric design

  11. Severe accident analysis to prevent high pressure scenarios in the EPR TM

    International Nuclear Information System (INIS)

    Azarian, G.; Gandrille, P.; Gasperini, M.; Klein, R.

    2010-01-01

    The EPR TM has incorporated several design features in order to specifically address major severe accident safety issues. In particular, it was designed with the objective to transfer high pressure core melt scenarios into a low pressure scenario with high reliability so that a high pressure vessel failure can be practically eliminated. It is the key issue in the defense-in-depth approach, for a postulated severe accident with core melting, to prevent any risk of containment failure due to possible Direct Containment Heating or due to reactor vessel rocketing which results from vessel failure at high pressure. Temperature-induced steam generator tube rupture, which could lead to a radiological containment bypass, has also to be prevented. On the basis of the analysis of the main high pressure core melt scenarios which are calculated with the MAAP4.07 code which was developed to support the EPR TM, this paper explores the benefits of primary depressurization by dedicated valves on transient evolutions. It specifically addresses the thermal response of the structures by sensitivity studies involving the timing of valve actuation. It outlines that a grace period of at least one hour is available for a delayed valve actuation without inducing excessive loads and without increasing the risk of a temperature-induced steam generator tube rupture. (authors)

  12. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  13. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. 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

    International Nuclear Information System (INIS)

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

    2017-01-01

    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.

  15. Core to surge-line energy transport in a severe accident scenario

    International Nuclear Information System (INIS)

    Marzo, M. di; Almenas, K.; Gopalnarayanan, S.

    1994-01-01

    The analysis of loss of coolant accidents in a nuclear power plant, which progress to the stage where the core is uncovered, poses important safety related questions. One of these concerns the rate of energy transport to metal components of the primary system. An experimental program has been conducted at the Univ. of Maryland test facility which quantifies the rate of energy transfer from an uncovered core in a B ampersand W (once-through type steam generators) plant. SF 6 is used to simulate the natural circulation driving force of the high pressure steam expected at prototypical conditions. A time-dependent scaling methodology is developed to transpose experimental data to prototypical conditions. To achieve this transformation, a nominal fluid temperature increase rate of 1.0 degrees C/s is inferred from available TMI-2 event data. To bracket the range of potential prototypical transient scenarios, temperature ramps of 0.8 degrees C/s and 1.2 degrees C/s are also considered. Repeated tests, covering a range of test facility conditions, lead to estimated failure times at the surge line nozzle of 1.5 to 2 hours after initiation of the natural circulation phase of the transient

  16. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

  17. Radiological impact assessment for a severe accident scenario for a CANDU 6 type NPP

    International Nuclear Information System (INIS)

    Penescu, Maria; Mehedinteanu, Stefan; Cruceanu, Amalia; Ispas, Georgeta

    2004-01-01

    . The consequences of a given accidental release will depend on the values of a number of parameters, in particular, type and nature of the accident, the total amount of radioactive material and different radionuclides involved, the energy with which they are disposed in the environment, the nature of the surrounding environment and the mechanisms of radioactive dispersion and transfer. This paper presents some assessments on the LOCA type accident influence, without the intact loop cooling and ECCS in case of a hypothetical site. The developed assessments resulted in the necessity to evacuate the population up to a distance of about 10 km away from the sectors located in the wind blowing direction. Note that for a real situation of a site, smaller values of the doses are expected, considering other phenomena which may lead to the decrease of the dose values for individuals (the consideration of the meteorologic parameter hourly values, of precipitation, of the actual distribution of population, of the protection measure application, etc)

  18. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  19. The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

    2010-01-01

    KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

  20. Severe accidents in nuclear reactors

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

    2004-01-01

    The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

  1. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

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

    2001-01-01

    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......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

  2. Severe accident behavior

    International Nuclear Information System (INIS)

    Denning, R.S.

    1986-01-01

    The purpose of this paper is to provide an overview of severe accident behavior. The term source term is defined and a brief history of the regulatory use of source term is presented. The processes in severe accidents in light water reactors are described with particular emphasis on the relationships between accident thermal-hydraulics and chemistry. Those factors which have the greatest impact on predicted source terms are identified. Design differences between plants that affect source term estimation are also described. The principal unresolved issues are identified that are the focus of ongoing research and debate in the technical community

  3. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  4. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery management concentrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk, and goes further in considering and formulating the key issue: The most fruitful path to follow in reducing risk even further is through the planning of accident management

  5. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

    No, Young Gyu; Ahn, Kwang Il; Kim, Ju Hyun; Na, Man Gyun; Lim, Dong Hyuk

    2012-01-01

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  6. Monitoring severe accidents using AI techniques

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Institute of Nuclear Nonproliferation and Control, Daejon (Korea, Republic of)

    2012-05-15

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  7. Accident scenario diagnostics with neural networks

    International Nuclear Information System (INIS)

    Guo, Z.

    1992-01-01

    Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

  8. 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.

  9. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

    The events at Fukushima Daiichi have highlighted the importance of Severe Accident Management Guidelines (SAMGs). As the world has learned from the catastrophe and countries are considering changes to their nuclear regulatory programs, the content of SAMGs and their regulatory control are being evaluated. This presentation highlights several factors that are being addressed in the United States as rulemaking is underway pertaining to SAMGs. The question of how to be prepared for the unexpected is discussed with specific insights gleaned from Fukushima. (author)

  10. Scenarios catalog for the graphical console for analysis of severe accidents visualization of OEs, NAEs and calculation of source term of the NPP-LV

    International Nuclear Information System (INIS)

    Sandoval V, S.; Mendoza R, M. E.; Tijerina S, F.; Garcia C, T.

    2016-09-01

    A nuclear power plant is operated at all times within the design criteria of structures, systems and components, and according to the operation technical specifications. For different areas of work of a nuclear power plant is necessary to carry out practices in which is useful to have the prediction of the thermo-hydraulic and radiological progression of scenarios that imply exceeding that design bases, even reaching the damage of the fuel in different degree. During the exercises and drills of the External Plan of Emergency Response, the projection of doses is done to exercise the different tasks of the plan. To make the projection of doses is required to have the radiological source term of the scenario on which is practiced. Because of this, was identified the convenience of having a catalog of scenarios for which the radiological source term was calculated. In 2004, a first version of the catalog was produced for a power of 2027 MW, and in 2011 the catalog was updated for extended power conditions, 2317 MW. Both versions were made using the severe accident simulator MAAP-3B. That catalog consists of a form and an optical storage device. The form contains tables and figures in which the characteristics of the scenario to be practiced are searched and the electronic files of the corresponding radiological source term are located in the storage device. Due to the recent development of the graphical console for analysis of severe accidents, visualization of OEs, NAEs and calculation of the source term for the nuclear power plant of Laguna Verde (NPP-LV) CoGrAAS, the catalog printed was replaced by an electronic catalog for the CoGrAAS. The new catalog retains the philosophy of the previous catalog, constituted by a wide collection of scenarios that involve different circumstances and phenomena, that can be used to practice different tasks during training exercises or simulacrums, and combined with the following advantages: the scenario selection is made from an

  11. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  12. MELCOR 1.8.3 application to NUPEC M-7-1 test (ISP-35) and two hydrogen severe accident scenarios in a typical PWR plant

    International Nuclear Information System (INIS)

    Jimenez Garcia, M.A.; Martin-Fuertes, F.; Martin-Valdepenas, J.M.

    1997-01-01

    Combustion of the hydrogen released to the containment during a severe accident is one of the issues to establish the real threats to the third barrier integrity in nuclear power facilities. Computational efforts on management procedures, such as the containment spray operation, are being addressed at the CTN-UPM to cope with the problem. On top of this, studies about in-containment hydrogen distribution and combustion are currently carried out with the codes MELCOR 1.8.3 and ESTER 1.0-RALOC 2.2. In this study, MELCOR 1.8.3 has been validated against the NUPEC M-7-1 Test, which already showed in 1993 that a good agreement was reached out when the previous MELCOR 1.8.2 calculations were performed regarding to the helium distribution throughout the facility. Nevertheless, some discrepancies were detected when analysing wall and atmosphere temperatures. Generally, well-mixed atmosphere scenarios, in which the role played by the containment water spraying is of the major importance, appear when such a mechanism promotes the onset of convection driven flow patterns that rapidly homogenize the gas properties. The purpose of the new MELCOR 1.8.3 assessment is to take advantage of the newest implemented models to obtain a more realistic thermalhydraulics simulation. A variation case was also performed to highlight the influence of water spray operation. In a second part of the study, insights coming from the previous work were used to apply MELCOR 1.8.3 models to a SBO severe accident scenario management in a commercial 2700 MWt 3-loop W PWR containment

  13. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

    No, Young Gyu; Kim, Ju Hyun; Na, Man Gyun; Ahn, Kwang Il

    2011-01-01

    It is very difficult for nuclear power plant operators to monitor and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. The objective of this study is to develop and verify the monitoring for severe accidents using artificial intelligence (AI) techniques such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH) and fuzzy neural network (FNN). The SVC and PNN are used for event classification among the severe accidents. Also, GMDH and FNN are used to monitor for severe accidents. The inputs to AI techniques are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. In this study, 3 types of initiating events such as the hot-leg LOCA, the cold-leg LOCA and SGTR are considered and it is verified how well the proposed scenario identification algorithm using the GMDH and FNN models identifies the timings when the reactor core will be uncovered, when CET will exceed 1200 .deg. F and when the reactor vessel will fail. In cases that an initiating event develops into a severe accident, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful for operators to perform severe accident management

  14. Risk assessment of complex accident scenarios

    International Nuclear Information System (INIS)

    Kluegel, Jens-Uwe

    2012-01-01

    The use of methods of risk assessment in accidents in nuclear plants is based on an old tradition. The first consistent systematic study is considered to be the Rasmussen Study of the U.S. Nuclear Regulatory Commission, NRC, WASH-1400. Above and beyond the realm of nuclear technology, there is an extensive range of accident, risk and reliability research into technical-administrative systems. In the past, it has been this area of research which has led to the development of concepts of safety precautions of the type also introduced into nuclear technology (barrier concept, defense in depth, single-failure criterion), where they are now taken for granted as trivial concepts. Also for risk analysis, nuclear technology made use of methods (such as event and fault tree analyses) whose origins were outside the nuclear field. One area in which the use of traditional methods of probabilistic safety analysis is encountering practical problems is risk assessment of complex accident scenarios in nuclear technology. A definition is offered of the term 'complex accident scenarios' in nuclear technology. A number of problems are addressed which arise in the use of traditional PSA procedures in risk assessment of complex accident scenarios. Cases of complex accident scenarios are presented to demonstrate methods of risk assessment which allow robust results to be obtained even when traditional techniques of risk analysis are maintained as a matter of principle. These methods are based on the use of conditional risk metrics. (orig.)

  15. 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

  16. Severe accident recriticality analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. E-mail: wiktor.frid@ski.se; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H

    2001-11-01

    Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, 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 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{sup -1} injected into the downcomer. Since 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{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

  17. Severe accident recriticality analyses (SARA)

    International Nuclear Information System (INIS)

    Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H.

    2001-01-01

    Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, 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 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, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

  18. Severe Accident Recriticality Analyses (SARA)

    International Nuclear Information System (INIS)

    Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

    1999-11-01

    Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B 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

  19. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  20. Sarnet lecture notes on nuclear reactor severe accident phenomenology

    International Nuclear Information System (INIS)

    Trambauer, K.; Adroguer, B.; Fichot, F.; Muller, C.; Meyer, L.; Breitung, W.; Magallon, D.; Journeau, C.; Alsmeyer, H.; Housiadas, C.; Clement, B.; Ang, M.L.; Chaumont, B.; Ivanov, I.; Marguet, S.; Van Dorsselaere, J.P.; Fleurot, J.; Giordano, P.; Cranga, M.

    2008-01-01

    The 'Severe Accident Phenomenology Short Course' is part of the Excellence Spreading activities of the European Severe Accident Research NETwork of Excellence SARNET (project of the EURATOM 6. Framework programme). It was held at Cadarache, 9-13 January 2006. The course was divided in 14 lectures covering all aspects of severe accident phenomena that occur during a scenario. It also included lectures on PSA-2, Safety Assessment and design measures in new LWR plants for severe accident mitigation (SAM). This book presents the lecture notes of the Severe Accident Phenomenology Short Course and condenses the essential knowledge on severe accident phenomenology in 2008. (authors)

  1. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2013-10-01

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  2. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  3. Radiological accidents, scenarios, planning and answers

    International Nuclear Information System (INIS)

    Solis Delgado, Alexander.

    2008-01-01

    Radiological accidents, scenarios and the importance of a good planning to prevent and control these types of accidents are presented. The radiation can be only one of the risks in an accident, most of dominant radiological risks are not radiological (fire, toxic gases, etc.). The common causes of radiological accidents, potential risks such as external irradiation, internal contamination and the environment pollution are highlighted. In addition, why accidents happen and how they evolve is explained. It describes some incidents with the radiation occurred in Costa Rica from 1993 to 2007. The coordination of emergency management in Costa Rica in relation to a radiological accident, and some mechanisms of action that have practiced in other places are focuses. Among the final considerations are the need to finalize the national plan for radiological emergencies as a tool of empowerment for the teams of emergency care and the availability of information. Likewise the processes of communication, coordination and cooperation to avoid chaos, confusion and crisis are also highlighted [es

  4. Development of severe accident management advisory and training simulator (SAMAT)

    International Nuclear Information System (INIS)

    Jeong, K.-S.; Kim, K.-R.; Jung, W.-D.; Ha, J.-J.

    2002-01-01

    The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management

  5. A Study on the Coupled FEM-Analysis for Reactor Vessel Lower Head of APR1400 under the Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyonam; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    For the stabilization of the RPV the in-vessel retention strategy with external reactor vessel cooling (IVR-ERVC) is adopted in APR1400. Under this severe accident condition, a good understanding of the mechanical behavior of the reactor vessel lower head (RVLH) is necessary both for verification of structural integrity and for improving the design applying appropriate accident mitigation strategies. The purpose of this study is to develop the analysis method of the RVLH with thermo-mechanical analysis using FEM tool (ANSYS v.15) in case of core-melting severe accident condition, and then analyze the RVLH of APR1400 including creep behavior. The plastic strain can be the major cause of lower head failure on the reactor vessel, and the creep cannot be not negligible factor of the failure under the severe accident condition. In the study, we applied constant convection coefficient at assumed temperature on the outside wall of RPV and substitute creep data of SA-508. In addition, it was found that the steel ablation at the interface between corium and vessel steel is not only a thermal phenomenon in the METCOR experiments. Corrosion processes and the formation of eutectics lead to the erosion of the vessel steel at temperatures that are significantly lower than the melting temperature of steel. It called thermo-chemical attack of the corium (corrosion). Reduced wall thickness because of the thermo-chemical effect by corium increase the equivalent plastic strain, and decrease the minimum time to reach 20% creep strain.

  6. Jose Cabrera NPP severe accident management activities

    International Nuclear Information System (INIS)

    Blanco, J.; Almeida, P.; Saiz, J.; Sastre, J.L.; Delgado, R.

    1998-01-01

    To prepare a common acting plan with respect to Severe Accident Management, in 1994 was founded the severe accident management ''ad-hoc'' working group from the Spanish Westinghouse PWR Nuclear Power Plant Owners Group. In this group actively collaborated the Jose Cabrera NPP Training Centre and the Department of Nuclear Engineering of UNION FENOSA. From this moment, Jose Cabrera NPP began the planning of its specific Severe Accident Management Program, which main point are Severe Accident Management Guidelines (SAMG). To elaborate this guidelines, the Spanish translation of Westinghouse Owners Group (WOG) Severe Accident Management Guidelines were considered the reference documents. The implementation of this Guidelines to Jose Cabrera NPP started on January 1997. Once the specific guidelines have been implemented to the plant, training activities for the personnel involved in severe accident issues will be developed. To prepare the training exercises MAAP4 code will be used, and with this intention, a specific Jose Cabrera NPP MAAP-GRAAPH screen has been developed. Furthermore, a wide selection of MAAP input files for the simulation of different scenarios and accidental events is available. (Author)

  7. Interaction of radionuclides in severe accident conditions

    International Nuclear Information System (INIS)

    Nagrale, Dhanesh B.; Bera, Subrata; Deo, Anuj Kumar; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Nuclear power plants are designed with inherent engineering safety systems and associated operational procedures that provide an in-depth defence against accidents. Radionuclides such as Iodine, Cesium, Tellurium, Barium, Strontium, Rubidium, Molybdenum and many others may get released during a severe accident. Among these, Iodine, one of the fission products, behaviour is significant for the analysis of severe accident consequences because iodine is a chemically more active to the potential components released to the environment. During severe accident, Iodine is released and transported in aqueous, organic and inorganic forms. Iodine release from fuel, iodine transport in primary coolant system, containment, and reaction with control rods are some of the important phases in a severe accident scenario. The behaviour of iodine is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. The presence of gaseous organic compounds and oxidizing compounds on iodine, reactions of aerosol iodine with boron and formation of cesium iodide which results in more volatile iodine release in containment play significant roles. Water radiolysis products due to presence of dissolved impurities, chloride ions, organic impurities should be considered while calculating iodine release. Containment filtered venting system (CFVS) consists of venturi scrubber and a scrubber tank which is dosed with NaOH and NaS_2O_3 in water where iodine will react with the chemicals and convert into NaI and Na_2SO_4. This paper elaborates the issues with respect to interaction of radionuclides and its consideration in modeling of severe accident. (author)

  8. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

    Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

    2014-01-01

    Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

  9. Vaporization of structural materials in severe accidents

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1982-01-01

    Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten pool. In many accident scenarios, the coefficient can be calculated from existing correlations for mass transfer by natural convection. Data from SASCHA fuel melting tests (Karlsruhe, Germany) show that the partial pressures of many of the melt components (Fe, Cr, Co, Mn, Sn) required for the model can be calculated from the vapor pressures of the pure species and Raoult's law. These calculations indicate much lower aerosol concentrations than reported in previous studies

  10. A CANDU Severe Accident Analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

    2006-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

  11. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

    2014-01-01

    The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

  12. Accident scenarios triggered by lightning strike on atmospheric storage tanks

    International Nuclear Information System (INIS)

    Necci, Amos; Argenti, Francesca; Landucci, Gabriele; Cozzani, Valerio

    2014-01-01

    Severe Natech accidents may be triggered by lightning strike affecting storage tanks containing relevant inventories of hazardous materials. The present study focused on the identification of event sequences and accident scenarios following lightning impact on atmospheric tanks. Reference event trees, validated using past accident analysis, are provided to describe the specific accident chains identified, accounting for reference protection and mitigation safety barriers usually adopted in current industrial practice. An overall methodology was outlined to allow the calculation of the expected frequencies of final scenarios following lightning impact on atmospheric storage tanks, taking into account the expected performance of available safety barriers. The methodology was applied to a case study in order to better understand the data that may be obtained and their importance in the framework of quantitative risk assessment (QRA) and of the risk management of industrial facilities with respect to external hazards due to natural events. - Highlights: • Event sequences following lightning impact on atmospheric tanks were identified. • Reference event trees including standard safety barriers were obtained. • Safety barriers applied in industrial practice were assessed to quantify event trees. • Frequencies of final scenarios following lightning impact on tanks were calculated. • Natech scenarios caused by lightning have an important influence on risk profiles

  13. Robot dispatching Scenario for Accident Condition Monitoring of NPP

    International Nuclear Information System (INIS)

    Kim, Jongseog

    2013-01-01

    In March of 2011, unanticipated big size of tsunami attacks Fukushima NPP, this accident results in explosion of containment building. Tokyo electric power of Japan couldn't dispatch a robot for monitoring of containment inside. USA Packbot robot used for desert war in Iraq was supplied to Fukushima NPP for monitoring of high radiation area. Packbot also couldn't reach deep inside of Fukushima NPP due to short length of power cable. Japanese robot 'Queens' also failed to complete a mission due to communication problem between robot and operator. I think major reason of these robot failures is absence of robot dispatching scenario. If there was a scenario and a rehearsal for monitoring during or after accident, these unanticipated obstacles could be overcome. Robot dispatching scenario studied for accident of nuclear power plant was described herein. Study on scenario of robot dispatching is performed. Flying robot is regarded as good choice for accident monitoring. Walking robot with arm equipped is good for emergency valve close. Short time work and shift work by several robots can be a solution for high radiation area. Thin and soft cable with rolling reel can be a good solution for long time work and good communication

  14. Prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Weisshaeupl, H.

    1996-01-01

    For the European Pressurized water Reactor (EPR), jointly developed by French and German industry, great emphasis is laid to gain further improvement in prevention of severe accidents based on the accumulative experience and proven technology of the French and German PWR reactors. In this evolutionary development, a balanced and comprehensive approach in respect to implement new passive features has been chosen. Improvements in each step of the defense in depth concept lead to a further decrease in the probability of occurrence of a severe accident with partial or even gross melting of the core. The different phenomenons that occur during such an hypothetical accident must be taken into account during the conception of specific measurements necessary to mitigate accident consequences. To cope with the consequences of a severe accident with core melt down means to deal with different phenomena which may threaten the integrity of the containment or may lead to an enhanced fission product release into the environment: high pressure reactor pressure vessel failure; energetic molten fuel coolant interaction; direct containment heating, molten core concrete interaction; hydrogen combustion; long term pressure and temperature increase in the containment. The EPR approach follows the recommendations from the DFD (Deutsch-Franzosischer Direktionsausschuss), jointly prepared by the French and German safety authorities. The EPR concept consist to prevent or eliminate as far as possible scenarios which are connected with high loads (high pressure failure of the reactor pressure vessel, or global hydrogen detonation etc..) by dedicated design provisions, and to deal with the consequences of severe accident scenarios which are not ruled out by specific safety measures. The measures comprise: the primary system depressurization; the control of hydrogen; the stabilisation and cooling of the melted core; the containment heat removal. They are completed by specific characteristics

  15. Overview of severe accident research at JAERI

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1999-01-01

    Severe accident research at JAERI aims at the confirmation of the safety margin, the quantification of the associated risk, and the evaluation of the effectiveness of the accident management measures of the nuclear power reactors, in accordance with the government five-year nuclear safety research program. JAERI has been conducting a wide range of severe accident research activities both in experiment and analysis, such as melt coolant interactions, fission product behaviors in coolant system, containment integrity and assessment of accident management measures. Molten core/coolant interaction and in-vessel molten coolability have been investigated in ALPHA Program. MUSE experiments in ALPHA Program has been conducted for the precise energy measurement due to steam explosion in melt jet and stratified geometries. In VEGA Program, which aims at FP release from irradiated fuels at high temperature and high pressure under various atmospheric conditions, the facility construction is almost completed. In WIND Program the revaporization of aerosols due to decay heating and also the integrity of the piping from this heat source are being investigated. Code development activities are in progress for an integrated source term analysis with THALES, fission product behaviors with ART, steam explosion with JASMINE, and in-vessel debris behaviors with CAMP. The experimental analyses and reactor application have made progress by participating international standard problem and code comparison exercises, along with the use of introduced codes, such as SCDAP/RELAP5 and MELCOR. The outcome of the severe accident research will be utilized for the evaluation of more reliable severe accident scenarios, detailed implementation of the accident management measures, and also for the future reactor development, basically through the sophisticated use of verified analytical tools. (author)

  16. Severe accident source term reassessment

    International Nuclear Information System (INIS)

    Hazzan, M.J.; Gardner, R.; Warman, E.A.; Jacobs, S.B.

    1987-01-01

    This paper summarizes the status of the reassessment of severe reactor accident source terms, which are defined as the quantity, type, and timing of fission product releases from such accidents. Concentration is on the major results and conclusions of analyses with modern methods for both pressurized water reactors (PWRs) and boiling water reactors (BWRs), and the special case of containment bypass. Some distinctions are drawn between analyses for PWRs and BWRs. In general, the more the matter is examined, the consequences, or probability of serious consequences, seem to be less. (author)

  17. 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.

  18. MELCOR DB Construction for the Severe Accident Analysis DB

    International Nuclear Information System (INIS)

    Song, Y. M.; Ahn, K. I.

    2011-01-01

    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

  19. The management of severe accidents

    International Nuclear Information System (INIS)

    Pelce, J.; Brignon, P.

    1987-01-01

    In considering severe accidents in water power reactors, a major problem that arises is how to manage them in such a way that the situation can be controlled as well as possible, from the aspects both of preventing serious damage to the core of limiting the discharge of radioactivity. A number of countries have announced provisions in the field of accident management, some already set up, others planned, but these mainly apply to preventing damage to the core. Part of this report deals with this aspect, to show that there is a fairly wide consensus on how problems should be approached. Attitudes vary, on the other hand, in the approach to mitigate radioactive release. In fact, few countries have proposed concrete steps to manage severe accidents in the final stages when the core is seriously damaged. Since it is difficult to compare different approaches, only the French approach is described. This description is however very brief, because in the five or six years since it was defined, the approach has been presented many times. The stress is placed more on the comments which this type of approach suggests, to make the subsequent general discussion easier

  20. Severe accident management. Prevention and Mitigation

    International Nuclear Information System (INIS)

    1992-01-01

    Effective planning for the management of severe accidents at nuclear power plants can produce both a reduction in the frequency of such accidents as well as the ability to mitigate their consequences if and when they should occur. This report provides an overview of accident management activities in OECD countries. It also presents the conclusions of a group of international experts regarding the development of accident management methods, the integration of accident management planning into reactor operations, and the benefits of accident management

  1. Complex accident scenarios modelled and analysed by Stochastic Petri Nets

    International Nuclear Information System (INIS)

    Nývlt, Ondřej; Haugen, Stein; Ferkl, Lukáš

    2015-01-01

    This paper is focused on the usage of Petri nets for an effective modelling and simulation of complicated accident scenarios, where an order of events can vary and some events may occur anywhere in an event chain. These cases are hardly manageable by traditional methods as event trees – e.g. one pivotal event must be often inserted several times into one branch of the tree. Our approach is based on Stochastic Petri Nets with Predicates and Assertions and on an idea, which comes from the area of Programmable Logic Controllers: an accidental scenario is described as a net of interconnected blocks, which represent parts of the scenario. So the scenario is firstly divided into parts, which are then modelled by Petri nets. Every block can be easily interconnected with other blocks by input/output variables to create complex ones. In the presented approach, every event or a part of a scenario is modelled only once, independently on a number of its occurrences in the scenario. The final model is much more transparent then the corresponding event tree. The method is shown in two case studies, where the advanced one contains a dynamic behavior. - Highlights: • Event & Fault trees have problems with scenarios where an order of events can vary. • Paper presents a method for modelling and analysis of dynamic accident scenarios. • The presented method is based on Petri nets. • The proposed method solves mentioned problems of traditional approaches. • The method is shown in two case studies: simple and advanced (with dynamic behavior)

  2. Improvement of the severe accident practice tool

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Takahashi, Shunsuke

    2016-01-01

    We developed the severe accident (SA) practice tool based on lessons learned in the accident at the Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Station. We utilized the developed SA practice tool and carried out the SA training for some employees of Kansai Electric Power Co., Inc. Afterwards, we examined the opinions given by trainees attending the training lecture and improved the SA practice tool to achieve a better educational effect. The main changes we made were improvement of the practice scenario for EAL judgments and addition of functions to the practice tool such as the EAL explanation document indication. As a result of having carried out the SA education using this practice tool, we determined the tool users could make the right EAL judgment and report the communication vote. Finally, we confirmed that the knowledge necessary for SA correspondence could be given satisfactorily by this practice tool. (author)

  3. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  4. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  5. 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)

  6. Relative evaluation on decommissioning accident scenarios of nuclear facilities

    International Nuclear Information System (INIS)

    Jeong, Kwan-Seong; Choi, Byung-Seon; Moon, Jei-Kwon; Hyun, Dong-Jun; Kim, Geun-Ho; Kim, Tae-Hyoung; Jo, Kyung-Hwa; Seo, Jae-Seok; Jeong, Seong-Young; Lee, Jung-Jun

    2012-01-01

    Highlights: ► This paper suggests relative importance on accident scenarios during decommissioning of nuclear facilities. ► The importance of scenarios can be performed by using AHP and Sugeno fuzzy method. ► The AHP and Sugeno fuzzy method guarantee reliability of the importance evaluation. -- Abstract: This paper suggests the evaluation method of relative importance on accident scenarios during decommissioning of nuclear facilities. The evaluation method consists of AHP method and Sugeno fuzzy integral method. This method will guarantee the reliability of relative importance evaluation for decommissioning accident scenarios.

  7. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1988-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  8. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

    Boesmans, B.; Auglaire, M.; Snoeck, J.

    1997-01-01

    The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

  9. 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.

  10. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

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

    2015-01-01

    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.

  11. Convective behaviour in severe accidents

    International Nuclear Information System (INIS)

    Clement, C.F.

    1988-01-01

    The nature and magnitude of the hazard from radioactivity posed by a possible nuclear accident depend strongly on convective behaviour within and immediately adjacent to the plant in question. This behaviour depends upon the nature of the vapour-gas-aerosol mixture concerned, and can show unusual properties such as 'upside-down' convection in which hot mixtures fall and cold mixtures rise. Predictions and criteria as to the types of behaviour which could possibly occur are summarised. Possible applications to present reactors are considered, and ways in which presently expected convection could be drastically modified are described. In some circumstances these could be used to suppress the radioactive source term or to switch its effect between distant dilute contamination and severe local contamination. (author). 8 refs, 2 figs, 2 tabs

  12. Severe accidents: in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    A ''severe'' nuclear accident refers to a reactor accident that could exceed reactor design specifications to such a degree as to prevent cooling of the reactor's core by normal means. This report summarizes the work of a NEA Senior Group of Experts who have studied the potential response of existing light-water reactors to severe accidents and have found that current designs of reactors are far more capable of coping with severe accidents than design specifications would suggest. The report emphasises the specific knowledge and means that can be used for diagnosing a severe accident and for managing its progression in order to prevent or mitigate its consequences

  13. SAMEX: A severe accident management support expert

    International Nuclear Information System (INIS)

    Park, Soo-Yong; Ahn, Kwang-Il

    2010-01-01

    A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R and D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.

  14. System 80+ design features for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

    1993-01-01

    ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

  15. Development of a severe accident training simulator using a MELCOR code

    International Nuclear Information System (INIS)

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

    2002-03-01

    Nuclear power plants' severe accidents are, despite of their rareness, very important in safety aspects, because of their huge damages when occurred. For the appropriate execution of severe accident strategy, more information for decision-making are required because of the uncertainties included in severe accidents. Earlier NRC raised concerns over severe accident training in the report NUREC/CR-477, and accordingly, developing effective training tools for severe accident were emphasized. In fact the training tools were requested from industrial area, nevertheless, few training tools were developed due to the uncertainties in severe accidents, lacks of analysis computer codes and technical limitations. SATS, the severe accident training simulator, is developed as a multi-purpose tools for severe accident training. SATS uses the calculation results of MELCOR, an integral severe accident analysis code, and with the help of SL-GMS graphic tools, provides dynamic displays of severe accident phenomena on the terminal of IBM PC. It aimed to have two main features: one is to provide graphic displays to represent severe accident phenomena and the other is to process and simulate severe accident strategy given by plant operators and TSC staffs. Severe accident strategies are basically composed of series of operations of available pumps, valves and other equipments. Whenever executing strategies with SATS, the trainee should follow the HyperKAMG, the on line version of the recently developed severe accident guidance (KAMG). Severe accident strategies are closely related to accidents scenarios. TLOFW and LOCA , two representative severe accident scenarios of Uljin 3,4, are developed as a built-in scenarios of SATS. Although SATS has some minor problems at this time, we expect SATS will be a good severe accident training tool after the appropriate addition of accident scenarios. Moreover, we also expect SATS will be a good advisory tool for the severe accident research

  16. Cost per severe accident as an index for severe accident consequence assessment and its applications

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2014-01-01

    The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100 MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. - Highlights: • Cost per severe accident is used for severe accident consequence assessment. • Assessments of health, economic, social and environmental impacts are included. • Radiation effect, relocation and decontamination costs are important cost components. • Cost per severe accident can be used to optimize radiation protection measures. • Effects of accident management can be estimated using the cost per severe accident

  17. Severe accident research in France

    International Nuclear Information System (INIS)

    Duco, J.; Reocreux, M.; Tattegrain, A.

    1988-01-01

    French PWR power plant design relies basically on a deterministic approach. Nevertheless, an overall safety objective was issued in 1977 by the safety authority which set an upper probability limit for having unacceptable consequences; this resulted, in particular, in the elaboration of the ''H'' procedures, aimed at reducing significantly the risk of core uncovery subsequent to the loss of redunbant safety-related systems. The U1 symptom-oriented procedure, based on the nuclear steam supply system ''cooling states'', was introduced later, in order to prevent core melting in situations where the operating crew was confused by multiple failures and/or inappropriate previous actions. In the event that a core-melt should occur, the ultimate procedures U2, U4 and U5 - the latter providing a venting of the containment through a filtration system - should enable the radioactive releases to be limited to characteristics compatible with the feasibility of the off-site emergency plans. Such emergency management procedures necessitate a significant study effort in order to be elaborated and qualified; this also presupposes that an adequate level of scientific knowledge has been gained as regards the response of specific components of a PWR under beyond-design conditions. The purpose of severe accident research in France is to attain a level of basic knowledge such that emergency procedures may be conceived and ultimately tested

  18. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

    2002-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

  19. Severe accident training simulator APROS SA

    International Nuclear Information System (INIS)

    Raiko, Eerikki; Salminen, Kai; Lundstroem, Petra; Harti, Mika; Routamo, Tomi

    2003-01-01

    APROS SA is a severe accident training simulator based on the APROS simulation environment. APROS SA has been developed in Fortum Nuclear Services Ltd to serve as a training tool for the personnel of the Loviisa NPP. Training with APROS SA gives the personnel a deeper understanding of the severe accident phenomena and thus it is an important part of the implementation of the severe accident management strategy. APROS SA consists of two parts, a comprehensive Loviisa plant model and an external severe accident model. The external model is an extension to the Loviisa plant model, which allows the simulation to proceed into the severe accident phase. The severe accident model has three submodels: the core melting and relocation model, corium pool model and fission product model. In addition to these, a new thermal-hydraulic solver is introduced to the core region of the Loviisa plant model to replace the more limited APROS thermal-hydraulic solver. The full APROS SA training simulator has a graphical user interface with visualizations of both severe accident management panels at the operator room and the important physical phenomena during the accident. This paper describes the background of the APROS SA training simulator, the severe accident submodels and the graphical user interface. A short description how APROS SA will be used as a training tool at the Loviisa NPP is also given

  20. 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

  1. Identification of reference accident scenarios in SEVESO establishments

    International Nuclear Information System (INIS)

    Delvosalle, C.; Fievez, C.; Pipart, A.; Fabrega, J. Casal; Planas, E.; Christou, M.; Mushtaq, F.

    2005-01-01

    In the frame of the ESREL special session on ARAMIS project, this paper aims at presenting the work carried out in the first Work Package, devoted to the definition of accident scenarios. This topic is a key-point in risk assessment, and serves as basis for the whole risk quantification. A first part of the work aims at building a Methodology for the Identification of Major Accident Hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. This work is coupled with an historical analysis of accidents. In a second part, influence of safety devices and policies will be considered, in order to build a Methodology for the Identification of Reference Accident Scenarios (MIRAS). This last one will take into account safety systems and lead to obtain more realistic scenarios

  2. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

    1992-12-01

    In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

  3. Severe accident analysis and management in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

    Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

  4. Neural network-based expert system for severe accident management

    International Nuclear Information System (INIS)

    Klopp, G.T.; Silverman, E.B.

    1992-01-01

    This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway at Commonwealth Edison Company (CECo). Phase I successfully demonstrated the feasibility of Artificial Neural Networks to support several of the objectives of severe accident management. Simulated accident scenarios were generated by the Modular Accident Analysis Program (MAAP) code currently in use by CECo as part of their Individual Plant Evaluations (IPE)/Accident Management Program. The primary objectives of the second phase were to develop and demonstrate four capabilities of neural networks with respect to nuclear power plant severe accident monitoring and prediction. The results of this work would form the foundation of a demonstration system which included expert system performance features. These capabilities included the ability to: (1) Predict the time available prior to support plate (and reactor vessel) failure; (2) Calculate the time remaining until recovery actions were too late to prevent core damage; (3) Predict future parameter values of each of the MAAP parameter variables; and (4) Detect simulated sensor failure and provide best-value estimates for further processing in the presence of a sensor failure. A variety of accident scenarios for the Zion and Dresden plants were used to train and test the neural network expert system. These included large and small break LOCAs as well as a range of transient events. 3 refs., 1 fig., 1 tab

  5. National practices in relation to severe accidents

    International Nuclear Information System (INIS)

    Soda, Kunihisa

    1989-01-01

    After the accidents at Three Mile Island and Chernobyl, many studies have been carried out on severe accidents by various organizations including IAEA and OECD/CSNI. In the present article, measures taken in different countries against severe accidents are outlined based on the results of these studies. In Sweden, policies for the management of a severe accident and reduction in the release of radioactive materials were established based on reports issued by the Atomic Energy Committee, which was set up after the Three Mile Island accident. The current policies require that filter vents be provided where necessary. France, following Sweden, adopted the use of filter vents. Operation procedures to be followed in the event of a severe accident have been established in the nation. The measures against severe accidents adopted in West Germany mainly focus on the weakening of the effects of accidents, and are not covered by the design standards. The use of filter vents are also required in Finland and Switzerland. In the U.S., a program for individual plant examination will be implemented over the three-year period beginning in 1989. Studies on measures against severe accidents seem to be performed also in the Soviet Union. (N.K.)

  6. 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.

  7. Monitoring and operation system for severe accidents

    International Nuclear Information System (INIS)

    Fukui, Toshiki; Niida, Shinji; Kato, Yumeto

    2017-01-01

    Monitoring and operation system for Severe Accidents (SA-MOS) is a compact Instrumentation and Control (I and C) system developed by Mitsubishi Heavy Industries (MHI) and certificated by the Japanese Nuclear Regulatory Agency (NRA) as a design application for Japanese existing PWR nuclear power plants. The system is tailored to provide monitoring and operation for Severe Accident (SA) conditions, and consists of digitalized I and C System, Human Systems Interface (HSI) system and Power Supply (PS) system as further improvement of reliability and safety. This design plans to be applied to the next Japanese PWR plants. In accordance with the new regulatory standards that NRA has established corresponding to the Fukushima accident, a long-term Station Black Out (SBO) scenario and 24-hours power supply by the storage battery in case of SA has been required. In order to address 24-hours power supply requirement in SA condition, the storage battery volume shall be increased. However, it may be difficult to introduce additional batteries to the existing plant site because of room space constraints, etc. Therefore, power distributions for the facilities which are only used for Design Basis Accident (DBA), are shut down in order to secure 24-hours operations of facilities for SA conditions including SA-MOS. That enables efficient battery resource operations as well as optimizes room space factors shared by battery cabinets. Another benefit is to introduce dedicate HSI system for SA condition and operators shift their operations using that dedicated HSI system to cope with SA events. That can reduce operator workload which forces operators to verify or choose which controllers and indicators are available in SA conditions. Furthermore, application of SA-MOS, secures the independence of the layers (DBA⇔SA) as well as secures the plant data transfer for SA conditions outside of plant. Those plant data assets can be shared by plant operation supporting personnel and

  8. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Valle Cepero, R.; Castillo Alvarez, J.; Ramon Fuente, J.

    1996-01-01

    For the assessment of the safety of nuclear power plants it is of great importance the analyses of severe accidents since they allow to estimate the possible failure models of the containment, and also permit knowing the magnitude and composition of the radioactive material that would be released to the environment in case of an accident upon population and the environment. This paper presents in general terms the basic principles for conducting the analysis of severe accidents, the fundamental sources in the generation of radionuclides and aerosols, the transportation and deposition processes, and also makes reference to de main codes used in the modulation of severe accidents. The final part of the paper contents information on how severe accidents are dialed with the regulatory point view in different countries

  9. Chemical phenomena under severe accident conditions

    International Nuclear Information System (INIS)

    Powers, D.A.

    1988-01-01

    A severe nuclear reactor accident is expected to involve a vast number of chemical processes. The chemical processes of major safety significance begin with the production of hydrogen during steam oxidation of fuel cladding. Physico-chemical changes in the fuel and the vaporization of radionuclides during reactor accidents have captured much of the attention of the safety community in recent years. Protracted chemical interactions of core debris with structural concrete mark the conclusion of dynamic events in a severe accident. An overview of the current understanding of chemical processes in severe reactor accident is provided in this paper. It is shown that most of this understanding has come from application of findings from other fields though a few areas have in the past been subject to in-depth study of a fundamental nature. Challenges in the study of severe accident chemistry are delineated

  10. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

    Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

    2014-01-01

    Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

  11. Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

    International Nuclear Information System (INIS)

    Lee, Yongjin; Jae, Moosung

    2014-01-01

    The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

  12. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

  13. Use of simulators in severe accident management

    International Nuclear Information System (INIS)

    Evans, R.C.

    1994-01-01

    The U.S. nuclear utility industry is moving in a deliberate fashion through a coordinated industry severe accident working group to study and augment, where appropriate, the existing utility organizational and emergency planning structure to address accident and severe accident management. Full-scope simulators are used extensively to train licensed operators for their initial license examinations and continually thereafter in licensed operator requalification training and yearly examinations. The goal of the training (both initial and requalification) is to ensure that operators possess adequate knowledge, skills and abilities to prevent an event from progressing to core damage. The use of full-scope simulators in severe accident management training is in large part viewed by the industry as being premature. The working group study has not progressed to the point where the decision to employ full-scope simulators can be logically considered. It is not however premature to consider part-task or work station simulators as invaluable research tools to support the industry's study. These simulators could be employed, subject to limitations in the current state of knowledge regarding severe accident progression and phenomenological responses, in the validation and verification (V and V) of severe accident models or codes as they are developed. The U.S. nuclear utility industry has made substantial strides in the past 12 years in the accident prevention, mitigation and management arena. These strides are a product of the industry's preference for a logical and systematic approach to change. (orig.)

  14. PWR severe accident mitigation measures, the french point of view

    International Nuclear Information System (INIS)

    Duco, J.; L'Homme, A.; Queniart, D.

    1990-01-01

    French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage, of the confinement of the radioactive substances in the containment. For a given scenario, one can almost always imagine a more severe scenario by postulating additional failures, but it is obvious that, as the severity of the imagined scenario increases, the probability of its occurrence tends towards zero. However, it does not appear reasonable to attempt to set a probability threshold below which the scenarios should be excluded. First of all, the higher the improbability of the scenarios, the greater the uncertainty in the calculation of their probability, with the result that the calculation is not very meaningful. Secondly, and more importantly, this approach ignores the essential problem of accident situation management. From the outset, French studies have been focused on controlling the development of these situations and mitigating their consequences by means of a series of appropriate actions involving, on the one hand, optimum use of the resources available in the installation during the course of the accident and, on the other hand, the taking of protective measures for the population. To attempt to prevent an initial event to degenerate into a severe accident leading to core meltdown if the proper actions are not taken, Electricite de France has proposed a new operating procedure based on the characterization of every possible cooling state of the core

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

    International Nuclear Information System (INIS)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee

    2016-01-01

    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

  16. 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.

  17. Deterministic analyses of severe accident issues

    International Nuclear Information System (INIS)

    Dua, S.S.; Moody, F.J.; Muralidharan, R.; Claassen, L.B.

    2004-01-01

    Severe accidents in light water reactors involve complex physical phenomena. In the past there has been a heavy reliance on simple assumptions regarding physical phenomena alongside of probability methods to evaluate risks associated with severe accidents. Recently GE has developed realistic methodologies that permit deterministic evaluations of severe accident progression and of some of the associated phenomena in the case of Boiling Water Reactors (BWRs). These deterministic analyses indicate that with appropriate system modifications, and operator actions, core damage can be prevented in most cases. Furthermore, in cases where core-melt is postulated, containment failure can either be prevented or significantly delayed to allow sufficient time for recovery actions to mitigate severe accidents

  18. MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

    International Nuclear Information System (INIS)

    Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

    2017-01-01

    The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

  19. Conclusions on severe accident research priorities

    International Nuclear Information System (INIS)

    Klein-Heßling, W.; Sonnenkalb, M.; Jacquemain, D.; Clément, B.; Raimond, E.; Dimmelmeier, H.; Azarian, G.; Ducros, G.; Journeau, C.; Herranz Puebla, L.E.; Schumm, A.; Miassoedov, A.; Kljenak, I.; Pascal, G.; Bechta, S.; Güntay, S.; Koch, M.K.; Ivanov, I.; Auvinen, A.; Lindholm, I.

    2014-01-01

    Highlights: • Estimation of research priorities related to severe accident phenomena. • Consideration of new topics, partly linked to the severe accidents at Fukushima. • Consideration of results of recent projects, e.g. SARNET, ASAMPSA2, OECD projects. - Abstract: The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II–III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency

  20. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    1989-11-01

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

  1. OSSA. A second generation of severe accident management

    International Nuclear Information System (INIS)

    Sauvage, E.C.; Musoyan, G.; Ducros, V.D.

    2009-01-01

    diagnostic directs the users to a series of system guidelines. The system approach allows a quick and general overview of the plant status. Following the course of an accident scenario evolving to a severe accident, the reader will be taken through the advanced concepts that are developed within the frame of the OSSA. Step by step, in place of the operators or the technical support group, he will experiment the OSSA. The selected severe accident scenario presented is a station blackout during which the late recovery will lead to severe accident. (author)

  2. SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

    International Nuclear Information System (INIS)

    Kymalainen, O.; Tuomisto, H.

    1997-01-01

    Some of the specific design features of IVO's Loviisa Plant, most notably the ice-condenser containment, strongly affect the plant response in a hypothetical core melt accident. They have together with the relatively stringent Finnish regulatory requirements forced IVO to develop a tailor made severe accident management strategy for Loviisa. The low design pressure of the ice-condenser containment complicates the design of the hydrogen management system. On the other hand, the ice-condensers and the water available from them are facilitating factors regarding in-vessel retention of corium by external cooling of reactor pressure vessel. This paper summarizes the Finnish severe accident requirements, IVO's approach to severe accidents, and its application to the Loviisa Plant

  3. A review of severe accident assessment

    International Nuclear Information System (INIS)

    Kawashima, Kei

    2000-01-01

    One of the most difficult problems on evaluation of external costs on nuclear power generation is value on a severe accident risk. Once forming a severe accident, its effect is very important and extends to a wide range, to give a lot of damages. It is a main area of study on externality of energy to compare various risks by means of price conversion at unit kWh. Here was outlined on research examples on main severe accident risks before then. A common fact on estimation cost such research examples is to limit it to direct cost (mainly to health damage) at accident phenomenon. As an actual problem, it is very difficult to substantially quantify such parameters because of basically belonging to social psychology. It is due to no finding out decisive evaluation method on this problem to be adopted conventional EED (Expert Expected Damages) approach in the ExternE Phase III, either. (G.K.)

  4. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent US Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in USNRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. 9 refs., 2 tabs

  5. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent U.S. Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in US-NRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. (author). 9 refs., 2 tabs

  6. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  7. United States position on severe accidents

    International Nuclear Information System (INIS)

    Ross, D.F.

    1988-01-01

    The United States policy on severe accidents was published in 1985 for both new plant applications and for existing plants. Implementation of this policy is in progress. This policy, aided by a related safety goal policy and by analysis capabilities emerging from improved understanding of accident phenomenology, is viewed as a logical development from the pioneering work in the WASH-1400 Reactor Safety Study published by the United States Nuclear Regulatory Commission (NRC) in 1975. This work provided an estimate of the probability and consequences of severe accidents which, prior to that time, had been mostly evaluated by somewhat arbitrary assumptions dating back 30 years. The early history of severe accident evaluation is briefly summarized for the period 1957-1979. Then, the galvanizing action of Three Mile Island Unit 2 (TMI-2) on severe accident analysis, experimentation and regulation is reviewed. Expressions of US policy in the form of rulemaking, severe accident policy, safety research, safety goal policy and court decisions (on adequacy of safety) are discussed. Finally, the NRC policy as of March 1988 is stated, along with a prospective look at the next few years. (author). 19 refs

  8. Uncertainties and severe-accident management

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1991-01-01

    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. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

  9. The development of severe accident analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heuy Dong; Cho, Sung Won; Kim, Sang Baek; Park, Jong Hwa; Lee, Kyu Jung; Park, Lae Joon; Hu, Hoh; Hong, Sung Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-07-01

    The objective of the development of severe accident analysis technology is to understand the severe accident phenomena such as core melt progression and to provide a reliable analytical tool to assess severe accidents in a nuclear power plant. Furthermore, establishment of the accident management strategies for the prevention/mitigation of severe accidents is also the purpose of this research. The study may be categorized into three areas. For the first area, two specific issues were reviewed to identify the further research direction, that is the natural circulation in the reactor coolant system and the fuel-coolant interaction as an in-vessel and an ex-vessel phenomenological study. For the second area, the MELCOR and the CONTAIN codes have been upgraded, and a validation calculation of the MELCOR has been performed for the PHEBUS-B9+ experiment. Finally, the experimental program has been established for the in-vessel and the ex-vessel severe accident phenomena with the in-pile test loop in KMRR and the integral containment test facilities, respectively. (Author).

  10. Severe accidents and terrorist threats at nuclear reactors

    International Nuclear Information System (INIS)

    Pollack, G.L.

    1987-01-01

    Some of the key areas of uncertainty are the nature of the physical and chemical interactions of released fission products and of the interactions between a molten core and concrete, the completeness and validity of the computer codes used to predict accidents, and the behavior of the containment. Because of these and other uncertainties, it is not yet possible to reliably predict the consequences of reactor accidents. It is known that for many accident scenarios, especially less severe ones or where the containment is not seriously compromised, the amount of radioactive material expected to escape the reactor is less, even much less, than was previously calculated. For such accidents, the predictions are easier and more reliable. With severe accidents, however, there is considerable uncertainty as to the predicted results. For accidents of the type that terrorists might cause - for example, where the sequence of failure would be unexpected or where redundant safety features are caused to fail together - the uncertainties are still larger. The conclusion, then, is that there are potential dangers to the public from terrorist actions at a nuclear reactor; however, because of the variety of potential terrorist threats and the incompleteness of the knowledge about the behavior of reactor components and fission products during accidents, the consequences cannot yet be assessed quantitatively

  11. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

    Duco, J.; Royen, J.; Rohde, J.; Frid, W.; De Boeck, B.

    1994-01-01

    The OECD Nuclear Energy Agency (NEA) organized in June 1994, in collaboration with the Swedish Nuclear Power Inspectorate (SKI), a Specialist Meeting on Selected Containment Severe Accident Management Strategies, to discuss their feasibility, effectiveness, benefits and drawbacks, and long-term impact. The meeting focused on water reactors, mainly on existing systems. The technical content covered topics such as general aspects of accident management strategies in OECD Member countries, hydrogen management techniques and other containment accident management strategies, surveillance and protection of the containment function. The main conclusions of the meeting are summarized in the paper. (author)

  12. Chemical considerations in severe accident analysis

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Swedish approach to information needs in severe accident situations

    Energy Technology Data Exchange (ETDEWEB)

    Soederman, E. (ES-Konsult AB, Stockholm (Sweden)); Karnik, P. (ES-Konsult AB, Stockholm (Sweden))

    1992-07-01

    In Sweden, systems for mitigating severe accidents have been installed at all plants and procedures have been implemented for accident management. This work has included the assessment of needs of information and the survivability of existing instrumentation during the various phases of an accident scenario. The approach has been pragmatic and based on existing knowledge of accident phenomenology and MAAP code calculations together with plant staff experience of detailed plant design and installation. During the early phases of accidents, which is defined to remain up to maximum fuel temperatures in the order of 800 C, the ordinary instrumentation is to a great extent useful. The reactor vessel level measurement is however identified to be weak in BWRs as soon as the core is partly uncovered. This has lead to the development of a Core Cooling Monitor. In later phases of accident scenarios, the general basis has been that no intrumentation inside the containment can survive. It has been analysed what information is strictly needed. It has been found that detailed information of the status inside the pressure vessel is of little importance after vessel penetration. Certain important information needs have been identified, that was not safely accessible from existing instrumentation. This had lead to complementary installations, using instruments inserted into the containment through protected guide tubes. Also for sampling of gas and water complementary installations have been made. (orig.)

  14. Swedish approach to information needs in severe accident situations

    International Nuclear Information System (INIS)

    Soederman, E.; Karnik, P.

    1992-01-01

    In Sweden, systems for mitigating severe accidents have been installed at all plants and procedures have been implemented for accident management. This work has included the assessment of needs of information and the survivability of existing instrumentation during the various phases of an accident scenario. The approach has been pragmatic and based on existing knowledge of accident phenomenology and MAAP code calculations together with plant staff experience of detailed plant design and installation. During the early phases of accidents, which is defined to remain up to maximum fuel temperatures in the order of 800 C, the ordinary instrumentation is to a great extent useful. The reactor vessel level measurement is however identified to be weak in BWRs as soon as the core is partly uncovered. This has lead to the development of a Core Cooling Monitor. In later phases of accident scenarios, the general basis has been that no intrumentation inside the containment can survive. It has been analysed what information is strictly needed. It has been found that detailed information of the status inside the pressure vessel is of little importance after vessel penetration. Certain important information needs have been identified, that was not safely accessible from existing instrumentation. This had lead to complementary installations, using instruments inserted into the containment through protected guide tubes. Also for sampling of gas and water complementary installations have been made. (orig.)

  15. A Scenario Proposal For A Radioactive Waste Transport Accident

    International Nuclear Information System (INIS)

    Salama, M.A.; Rashad, S.M.

    1999-01-01

    In spite of all precautions that being taken during radioactive materials transport accidents to ensure safe transportation of these materials; there is still a probability for accidents to occur which, may be accompanied by injury or death of persons and damage of property So, in response to the increasing possibilities of accidents in Egypt, the government had prepared an emergency response plan for radiological accidents to coordinate the response efforts of all the national agencies. Trends for use of the radioactive materials and sources inside the country for the purpose of medical diagnosis and treatment as well as in industrial applications, are increasing. The radioactive waste resulted from these activities are transported from the centres where these materials being used to the waste management facility where they are treated and finally disposed safely at disposal site. The aim of the emergency exercise scenario is to test not only the main components of the emergency plan but also the level of emergency preparedness; that is the effectiveness with which the actions or combined actions of the different organizations involved in an emergency can be put into practice. The motivation of the present paper was undertaken to give a scenario proposal for the radiological emergency actions taken in case of a transport accident for a radioactive waste material (type A- package ) transported by a vehicle from one of the medical centers to a disposal site, 40 km northeast of cairo

  16. NPP accident scenario. Which emergency measures are planned in Switzerland?

    International Nuclear Information System (INIS)

    Flury, Christoph

    2016-01-01

    As a consequence of the reactor accident in Fukushima the Swiss government has ordered an extensive analysis of emergency planning in case of a NPP accident Switzerland. A special working group has analyzed the possible improvements of Swiss emergency planning based on the experiences in Japan. Under the special direction of the Bundesamt fuer Bevoelkerungsschutz (BABS) the agreed improvements were integrated into the emergency concept. The reference scenarios have been re-assessed and the zone concept adapted. The emergency measures include shelter-type rooms (basement or window-less rooms), the preventive distribution of iodine pills, measures concerning agriculture, aquatic systems, preventive evacuation, traffic regulations, and delayed evacuation.

  17. 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)

    2017-08-01

    After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative would significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial 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. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.

  18. OSSA - An optimized approach to severe accident management: EPR application

    International Nuclear Information System (INIS)

    Sauvage, E. C.; Prior, R.; Coffey, K.; Mazurkiewicz, S. M.

    2006-01-01

    There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field

  19. Severe accident issue resolution -- definition and perspective

    International Nuclear Information System (INIS)

    Harper, F.T.

    1995-01-01

    The purpose of this discussion is to introduce the session on the Progress on the Resolution of Severe Accident Issues. There has been much work in the area of resolution of severe accident issues over the past few years. This work has been focused on those issues most important to risk as assessed by comprehensive studies such as NUREG-1150. In particular, issues associated with early containment failure have been analyzed. These efforts to resolve issues have been hampered by the fact that open-quotes issue resolutionclose quotes has not always been well defined. The term open-quotes issue resolutionclose quotes conjures tip different images for the regulator, the accident analyst, the physicist, and the probabalist. In fact it is common to have as many different images of issue resolution as there are people in the room. This issue is complicated by the fact that the uncertainty in severe accident issues is enormous. (When convolved, the quantitative uncertainty in an integrated analysis due to severe accident issues can span several orders of magnitude.) In this summary, hierarchy is presented in an attempt to add some perspective to the resolution of issues in the face of large uncertainties. Recommendations are also made for analysts communicating in the area of issue resolution

  20. [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.

  1. Computerized accident management support system: development for severe accident management

    International Nuclear Information System (INIS)

    Garcia, V.; Saiz, J.; Gomez, C.

    1998-01-01

    The activities involved in the international Halden Reactor Project (HRP), sponsored by the OECD, include the development of a Computerized Accident Management Support System (CAMS). The system was initially designed for its operation under normal conditions, operational transients and non severe accidents. Its purpose is to detect the plant status, analyzing the future evolution of the sequence (initially using the APROS simulation code) and the possible recovery and mitigation actions in case of an accident occurs. In order to widen the scope of CAMS to severe accident management issues, the integration of the MAAP code in the system has been proposed, as the contribution of the Spanish Electrical Sector to the project (with the coordination of DTN). To include this new capacity in CAMS is necessary to modify the system structure, including two new modules (Diagnosis and Adjustment). These modules are being developed currently for Pressurized Water Reactors and Boiling Water REactors, by the engineering of UNION FENOSA and IBERDROLA companies (respectively). This motion presents the characteristics of the new structure of the CAMS, as well as the general characteristics of the modules, developed by these companies in the framework of the Halden Reactor Project. (Author)

  2. Application of FFTBM to severe accidents

    International Nuclear Information System (INIS)

    Prosek, A.; Leskovar, M.

    2005-01-01

    In Europe an initiative for the reduction of uncertainties in severe accident safety issues was initiated. Generally, the error made in predicting plant behaviour is called uncertainty, while the discrepancies between measured and calculated trends related to experimental facilities are called the accuracy of the prediction. The purpose of the work is to assess the accuracy of the calculations of the severe accident International Standard Problem ISP-46 (Phebus FPT1), performed with two versions of MELCOR 1.8.5 for validation purposes. For the quantitative assessment of calculations the improved fast Fourier transform based method (FFTBM) was used with the capability to calculate time dependent code accuracy. In addition, a new measure for the indication of the time shift between the experimental and the calculated signal was proposed. The quantitative results obtained with FFTBM confirm the qualitative conclusions made during the Jozef Stefan Institute participation in ISP-46. In general good agreement of thermal-hydraulic variables and satisfactory agreement of total releases for most radionuclide classes was obtained. The quantitative FFTBM results showed that for the Phebus FPT1 severe accident experiment the accuracy of thermal-hydraulic variables calculated with the MELCOR severe accident code is close to the accuracy of thermal-hydraulic variables for design basis accident experiments calculated with best-estimate system codes. (author)

  3. Severe accident testing of a personnel airlock

    International Nuclear Information System (INIS)

    Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

    1988-01-01

    Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the U.S. Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800 degrees F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. The authors provide a detailed report on the test program

  4. Modelling and analysis of severe accidents for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the

  5. Heat transfer phenomena revelant to severe accidents

    International Nuclear Information System (INIS)

    Dallman, R.J.; Duffey, R.B.

    1990-01-01

    A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression

  6. Heat transfer phenomena relevant to severe accidents

    International Nuclear Information System (INIS)

    Dallman, R.J.; Duffey, R.B.

    1990-01-01

    A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression. 26 refs

  7. Thermal hydraulics of CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

  8. Core loss during a severe accident (COLOSS)

    International Nuclear Information System (INIS)

    Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

    2005-01-01

    The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

  9. Effect of RCIC Operating Conditions on the Accident Scenario in Fukushima Unit 2

    International Nuclear Information System (INIS)

    Kim, Sung Il; Park, Jong Hwa; Ha, Kwang Soon

    2015-01-01

    This study was conducted by using MELCOR 1.8.6. Fukushima unit 2 accident was analyzed using MELCOR in this study, and best estimate scenario with considering RCIC operating conditions was presented. Researches on the boiling water reactor (BWR) plant with reactor core isolation cooling (RCIC) system have been conducted. Research on the RCIC operation in Fukushima unit 2 was also conducted by Sandia National Laboratory. MELCOR analysis of the Fukushima unit 2 accident was conducted in the report and energy balance in wetwell was described by considering RCIC operation. However, the effect of RCIC operation condition on the accident scenario has not been studied. The operating conditions of RCIC system affect the pressures in wetwell and drywell, and the high pressure can make leakage path of fission product from PCV to reactor building. Thus it can be directly related with the amount of fission product which released to environment. In this study, severe accident on Fukushima unit 2 was analyzed considering the operating condition of RCIC system, and best estimated scenario was presented. In addition, the effect of RCIC turbine efficiency on the accident progression was examined. Energy balance in suppression chamber was also considered with discussion on the effect of torus room flooding level. It was found that the operating condition of RCIC turbine not only affects the variation of drywell pressure but also the amount of released fission products to environment. It was also confirmed that the RCIC turbine efficiency in the accident would be less than normal operating condition

  10. 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.

  11. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

    1998-01-01

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  12. Simulation of severe accidents in COTELS experiments

    International Nuclear Information System (INIS)

    Vasilev, Yu.S.; Zhdanov, V.S.; Kolodeshnikov, A.A.; Kadyrov, Kh. G.; Turkebaev, T.E.; Tsaj, K.V.; Suslov, E.E.

    1999-01-01

    At present, the issue of atomic reactor operation safety is of a great attention. It is evident that the accident accompanied with a core materials melting is an improbable event. To fully assess a hazard of a reactor use and enhance its safety, it is necessary to predict a possible accident progress and specify possible consequences of severe accidents and eliminating measures. In COTELS experiments, aimed at investigation of interaction of corium with concrete and water, the corium s imulator m elt is discharged on the concrete. The concrete erosion parameters, composition and rate of aerosol and gas escaping are recorded. The solidified melt and concrete fragments structure is studied after the testing, using the X-ray diffractometer DRON-3. This paper gives consideration to possible mechanisms of formation of uranium-containing and other phases of products of interaction of the corium melt with concrete and water

  13. Overview of severe accident research at KAERI

    International Nuclear Information System (INIS)

    Kim, H.D.; Kim, S.B.; Hong, S.W.; Kim, D.H.

    2000-01-01

    The severe accident research program at Korea Atomic Energy Research Institute, within the framework of governmental 10 year long-term nuclear R and D program, aims at the development of assessment techniques and accident management strategies for the prevention and mitigation of potential risk. The research program includes experimental efforts, development of phenomena specific models and development of an integrated computer code. The results of research program is intended to be utilized for the design of the advanced light water reactor and development of accident management strategies for the operating reactors. The main focused areas of recent investigation at KAERI are experiments on in-vessel core debris retention (SONATA-IV) and fuel coolant interaction (TROI) along with the development of models and integrated computer code (MIDAS). (author)

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

    International Nuclear Information System (INIS)

    Deng Jian; Cao Xuewu

    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. (authors)

  15. 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.

  16. An analysis of severe air transport accidents

    International Nuclear Information System (INIS)

    McClure, J.D.; Luna, R.E.

    1989-01-01

    The objective of this paper is to analyze the severity of aircraft accidents that may involve the air transport of radioactive materials (RAM). One of the basic aims of this paper is to provide a numerical description of the severity of aircraft transport accidents so that the accident severity can be compared with the accident performance standards that are specified in IAEA Safety Series 6, the international packaging standards for the safe movement of RAM. The existing packaging regulations in most countries embrace the packaging standards developed by the IAEA. Historically, the packaging standards for Type B packages have been independent of the transport mode. That is, if the shipment occurs in a certified packaging, then the shipment can take place by any transport mode. In 1975, a legislative action occurred in the US Congress which led to the development of a package designed specifically for the air transport of plutonium. Changes were subsequently made to the US packaging regulations in 10CFR71 to incorporate the plutonium air transport performance standards. These standards were used to certify the air transport package for plutonium which is commonly referred to as PAT-1 (US NRC). The PAT-1 was certified by the US Nuclear Regulatory Commission in September 1978

  17. Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

    2014-05-15

    700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

  18. Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

    International Nuclear Information System (INIS)

    McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

    1987-02-01

    This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

  19. Light water reactor severe accident seminar. Seminar presentation manual

    International Nuclear Information System (INIS)

    2004-01-01

    The topics covered in this manual on LWR severe accidents were: Evolution of Source Term Definition and Analysis, Current Position on Severe Accident Phenomena, Current Position on Fission Product Behavior, Overview of Software Models Used in Severe Accident Analysis, Overview of Plant Specific Source Terms and Their Impact on Risk, Current Applications of Severe Accident Analysis, and Future plans

  20. Light water reactor severe accident seminar. Seminar presentation manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The topics covered in this manual on LWR severe accidents were: Evolution of Source Term Definition and Analysis, Current Position on Severe Accident Phenomena, Current Position on Fission Product Behavior, Overview of Software Models Used in Severe Accident Analysis, Overview of Plant Specific Source Terms and Their Impact on Risk, Current Applications of Severe Accident Analysis, and Future plans.

  1. Assessment of uncertainties in severe accident management strategies

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Catton, I.; Dhir, V.K.; Okrent, D.

    1990-01-01

    Recent progress on the development of Probabilistic Risk Assessment (PRA) as a tool for qualifying nuclear reactor safety and on research devoted to severe accident phenomena has made severe accident management an achievable goal. Severe accident management strategies may involve operational changes, modification and/or addition of hardware, and institutional changes. In order to achieve the goal of managing severe accidents, a method for assessment of strategies must be developed which integrates PRA methodology and our current knowledge concerning severe accident phenomena, including uncertainty. The research project presented in this paper is aimed at delineating uncertainties in severe accident progression and their impact on severe accident management strategies

  2. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

    2014-01-01

    A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

  3. Development of Krsko Severe Accident Management Database (SAMD)

    International Nuclear Information System (INIS)

    Basic, I.; Kocnar, R.

    1996-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. Krsko Severe Accident Management Database documents the severe accident management activities which are developed in the NPP Krsko, based on the Krsko IPE (Individual Plant Examination) insights and Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidance). (author)

  4. Occupational Radiation Protection in Severe Accident Management

    International Nuclear Information System (INIS)

    2015-01-01

    As an early response to the Fukushima Daiichi NPP accident, the Information System on Occupational Exposure (ISOE) Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011: - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers /responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE official participants and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the-art ISOE report on best radiation protection management practices for proper radiation

  5. Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Faucett, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Haskin, Troy Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Luxat, Dave [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Geiger, Garrett [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Codella, Brittany [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-10-01

    Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

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

    International Nuclear Information System (INIS)

    Lee, NaRae; Lee, JinYong; Bang, YoungSuk; Lee, DooYong; Kim, HyeongTaek

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  8. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  9. Severe accidents and ESFR design issues

    International Nuclear Information System (INIS)

    Rineiski, A.

    2013-01-01

    Current SFR studies in Germany: ⇒ In support of European SFR studies, mainly on safety and safety-related (design optimization) issues; ⇒ ADS and SFR as main options for spent fuel management in studies on the possibility of P&T; ⇒ ESFR-type designs studied recently; ⇒ ASTRID-type designs to be studied in the future; ⇒ Particular area: modeling of severe accidents with SAS4A/SAS-SFR and SIMMER codes

  10. CANDU severe accident management guidance update

    International Nuclear Information System (INIS)

    Jones, L.; Popov, N.; Gilbert, L.; Weed, J.

    2014-01-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  11. CANDU severe accident management guidance update

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L., E-mail: lisa.m.jones@opg.com [Ontario Power Generation, Pickering, ON (Canada); Popov, N., E-mail: nik.popov@rogers.com [Candu Owners Group, Toronto, ON (Canada); Gilbert, L., E-mail: lovell.gilbert@brucepower.com [Bruce Power, Tiverton, ON (Canada); Weed, J., E-mail: jeff.weed@candu.gov [Candu Owners Group, Toronto, ON (Canada)

    2014-07-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  12. Development of Severe Accident Containment Analysis Package

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

  13. Nuclear power plant Severe Accident Research Plan

    International Nuclear Information System (INIS)

    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

  14. Method of assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems, and actions to prevent or mitigate a severe accident. A significant number of probabilistic safety assessments (PSAs) have been completed that yield the principal plant vulnerabilities. These vulnerabilities can be categorized as (1) dominant sequences with respect to core-melt frequency. (2) dominant sequences with respect to various risk measures. (3) dominant threats that challenge safety functions. (4) dominant threats with respect to failure of safety systems. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainties in key phenomena, operator behavior, system availability and behavior, and available information. This paper presents a methodology for assessing severe accident management strategies given the key uncertainties delineated at two workshops held at the University of California, Los Angeles. Based on decision trees and influence diagrams, the methodology is currently being applied to two case studies: cavity flooding in a pressurized water reactor (PWR) to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent vessel and/or containment failure

  15. Investigation of the different scenarios occurring in a PWR in case of a TMLB accident

    International Nuclear Information System (INIS)

    Pochard, R.; Dufresne, J.; Autrusson, B.

    1988-10-01

    Severe accidents in light water reactors fall into one of two main categories, depending on whether or not core meltdown is accompanied by a pressure buildup in the primary system. The way in which the accident develops is, in fact, largely conditioned by this pressure aspect: temperature distribution in the core and primary system resulting from natural convection gas streams; fuel clad failure mode, etc... One major effect of pressure buildup on the accident scenario is primary system failure under the combined actions of pressure and temperature. The purpose of the present paper is to present, after a detailed thermalhydraulic study, an analysis of the timing and location of the system failures in case of a TMLB accident on CPY french type reactor

  16. Application of ASTEC V2.0 to severe accident analyses for German KONVOI type reactors

    International Nuclear Information System (INIS)

    Nowack, H.; Erdmann, W.; Reinke, N.

    2011-01-01

    The integral code ASTEC is jointly developed by IRSN (Institut de Radioprotection et de Surete Nucleaire, France) and GRS (Germany). Its main objective is to simulate severe accident scenarios in PWRs from the initiating event up to the release of radioactive material into the environment. This paper describes the ASTEC modeling approach and the nodalisation of a KONVOI type PWR as an application example. Results from an integral severe accident study are presented and shortcomings as well as advantages are outlined. As a conclusion, the applicability of ASTEC V2.0 for deterministic severe accident analyses used for PSA level 2 and Severe Accident Management studies will be assessed. (author)

  17. Suppression Pools: paradigm of the thermalhydraulic effect on severe accidents

    International Nuclear Information System (INIS)

    Herranz, L. E.; Lopez del Pra, C.

    2016-01-01

    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)

  18. Control room habitability during severe accidents

    International Nuclear Information System (INIS)

    Siu, R.P.

    1989-01-01

    The requirements for protection of control room personnel against radiation hazards are specified in 10CFR50, Appendix A, GDC 19. The conventional approach involves a mechanistic evaluation of the radiation doses to control room personnel during design-basis accidents. In this study, an assessment of control room habitability during severe accidents is conducted. The potential levels of radiation hazards to control room personnel are evaluated in terms of both magnitude and probability of occurrence. The expected values for the probabilities of exceeding GDC-19 limits and the cumulative probability distributions of control room doses are determined. In this study, a pressurized water reactor with a large dry containment has been selected for analysis. The types of control rooms evaluated in this study include designs with: (a) filtered local intakes only, (b) filtered recirculation only, (c) filtered local intakes and recirculation, and (d) filtered dual remote intakes and recirculation. From the observations, it is concluded that, except for control room D, all other control room designs may require improvements in order to provide adequate radiation protection during severe accidents, particularly in terms of reducing whole-body gamma doses and skin doses. Potential design improvements include reduction of intake flows for concepts relying on pressurization, reduction in overall leakages, and control room pressurization through the use of bottled air supply

  19. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsen, S.; Walmod-Larsen, O.

    1986-08-01

    This report was commissioned by the Swedish State Power Board, who wanted a method for calculation of radiation doses in the surroundings of nuclear power plants caused by severe accidents. The PLUCON4 code were used for the calculations. A TC-SV-accident at Ringhals 1 wer chosen as example. A transient without shutdown leads to core meltdown through the reactor vessel. The pressure peak at the moment of vessel failure opens a safety valve in the dry well. Meteorolgical data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather were chosen Pasquill D with wind speed 8 m/s, and as extreme weather were chosen Pasquill F with wind speed 4.8 m/s. (author)

  20. Predicted occurrence rate of severe transportation accidents involving large casks

    International Nuclear Information System (INIS)

    Dennis, A.W.

    1978-01-01

    A summary of the results of an investigation of the severities of highway and railroad accidents as they relate to the shipment of large radioactive materials casks is discussed. The accident environments considered are fire, impact, crash, immersion, and puncture. For each of these environments, the accident severities and their predicted frequencies of occurrence are presented. These accident environments are presented in tabular and graphic form to allow the reader to evaluate the probabilities of occurrence of the accident parameter severities he selects

  1. Development of a totally integrated severe accident training system

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Park, Sun Hee; Choi, Young; Kim, Dong Ha

    2006-01-01

    Recently KAERI has developed the severe accident management guidance to establish the Korea standard severe accident management system. On the other hand the PC-based severe accident training simulator SATS has been developed, which uses the MELCOR code as the simulation engine. The simulator SATS graphically displays and simulates the severe accidents with interactive user commands. Especially the control capability of SATS could make a severe accident training course more interesting and effective. In this paper we will describe the development and functions of the electrical guidance module, HyperKAMG, and the SATS-HyperKAMG linkage system designed for a totally integrated and automated severe accident training. (author)

  2. Development of severe accident guidance module for the SATS simulator

    International Nuclear Information System (INIS)

    Kim, K.R.; Park, S.H.; Kim, D.H.; Song, Y.M.

    2004-01-01

    Recently KAERI has developed the severe accident management guidance to establish Korea standard severe accident management system. On the other hand the PC-based severe accident training simulator SATS has been developed, which uses MELCOR code as the simulation engine. SATS graphically displays and simulates the severe accidents with interactive user commands. The control capability of SATS could make severe accident training course more interesting and effective. In this paper we will describe the development and functions of the electrical hypertext guidance module HyperKAMG and the SATS-HyperKAMG linkage system for the severe accident management. (author)

  3. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-01-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions

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

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

    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.

  5. 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.

  6. Study Of Severe Accident Phenomena In Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sugiyanto; Antariksawan; Anhar, R.; Arifal

    2001-01-01

    Several phenomena that occurred in the light water reactor type of nuclear power plant during severe accident were studied. The study was carried out based on the results of severe accident researches in various countries. In general, severe accident phenomena can be classified into in-vessel phenomena, retention in the reactor coolant system, and ex-vessel phenomena. In-vessel retention has been recommended as a severe accident management strategy

  7. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  8. Possible consequences of severe accidents at the Lubiatowo site, Poland

    Science.gov (United States)

    Seibert, Petra; Philipp, Anne; Hofman, Radek; Gufler, Klaus; Sholly, Steven

    2014-05-01

    The construction of a nuclear power plant is under consideration in Poland. One of the sites under discussion is near Lubiatowo, located on the cost of the Baltic Sea northwest of Gdansk. An assessment of possible environmental consequences is carried out for 88 real meteorological cases with the Lagrangian particle dispersion model FLEXPART. Based on literature research, three reactor designs (ABWR, EPR, AP 1000) were identified as being under discussion in Poland. For each of the designs, a set of accident scenarios was evaluated and two source terms per reactor design were selected for analysis. One of the selected source terms was a relatively large release while the second one was a severe accident with an intact containment. Considered endpoints of the calculations are ground contamination with Cs-137 and time-integrated concentrations of I-131 in air as well as committed doses. They are evaluated on a grid of ca. 3 km mesh size covering eastern Central Europe.

  9. 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

  10. Severe Accidents in the Energy Sector

    International Nuclear Information System (INIS)

    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

  11. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  12. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  13. Several accidents about ERHRS of CEFR

    International Nuclear Information System (INIS)

    Zhang, D.

    2000-01-01

    An analysis of about several unusual accidents about Emergency Residual Heat Removal System (ERHRS) of China Experiment Fast Reactor (CEFR) is presented. CEFR is a pool-type sodium-cooled fast reactor. The ERHRS of this reactor is designed in passive principle, which enhances the interior reliability of CEFR. It consists of two sets of independent channels. Each channel is comprised of decay heat exchanger (DHX), intermediate circuit, sodium-air heat exchanger (AHX) and related auxiliary system. Both DHX are located in the hot pool of the main vessel directly, which is used to cool the hot sodium. The whole set of ERHRS is completely passive except the ventilation valves of AHX. But, as a very important set of engineered safety features which is the final way to remove the heat from the reactor core, it is necessary to pay attention to all of the possibilities that may reduce this ability. Several accidents are analyzed including when the ventilation valves couldn't be opened, when only one set of ERHRS could work and so on. The calculation results show that the ERHRS can keep the reactor in a safety status. Even though it is, experiments are still necessary in the view of engineering. (author)

  14. Benchmark accident scenarios for nuclear powered warship visits to Australian ports

    International Nuclear Information System (INIS)

    Frikken, A.J.

    1996-01-01

    Full text: Port safety arrangements for visits of nuclear powered warships (NPWs) to Australian ports require compliance with a number of Conditions of Entry. One of these Conditions of Entry is the provision of 'an operating safety organisation, competent to produce a suitable radiation monitoring program and able to initiate actions and provide services necessary to safeguard the public in the event of a release of radioactivity following an accident'. The States and Territories which receive NPW visits have the responsibility for this contingency planning, although the Commonwealth provides assistance through the Visiting Ships Panel Nuclear (VSP(N)). The visit of a NPW to an Australian port may only proceed if the VSP(N) is satisfied that the port safety plan has been exercised in sufficient depth to demonstrate its adequacy and efficacy. Emergency exercises are held on a regular basis in Australian ports which have been validated for visits by NPWs to ensure compliance with the Conditions of Entry. Important aspects of these exercises are the procedures for estimating radiation doses to members of the public following an accident, and the process for making decisions on the need for countermeasures based on the results of dose estimates. To exercise these aspects of the emergency response, detailed emergency exercise scenarios, including simulated radiological monitoring data, are required. To date, emergency exercises have usually been based on a severe and highly improbable scenario, termed the Reference Accident, which is used to assess the suitability of ports for visits by NPW's. The repeated adoption of this scenario does not adequately test the flexibility of the emergency plans to cope with all possible accidents, particularly more likely, less severe accidents. At the request of the VSP(N), the Nuclear Safety Bureau has analysed a spectrum of NPW accident scenarios and developed a set of Bench Mark Accident (BMA) scenarios for emergency response

  15. Use of PSA and severe accident assessment results for the accident management

    International Nuclear Information System (INIS)

    Jang, S. H.; Kim, H. G.; Jang, H. S.; Moon, S. K.; Park, J. U.

    1993-12-01

    The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management

  16. Use of PSA and severe accident assessment results for the accident management

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S H; Kim, H G; Jang, H S; Moon, S K; Park, J U [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    1993-12-15

    The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management.

  17. 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

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

    International Nuclear Information System (INIS)

    Park, S. Y.; Ahn, K. I.

    2015-01-01

    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

  19. Porosity effects during a severe accident

    International Nuclear Information System (INIS)

    Cazares R, R. I.; Espinosa P, G.; Vazquez R, A.

    2015-09-01

    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)

  20. 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)

  1. Severe accidents at nuclear power plants. Their risk assessment and accident management

    International Nuclear Information System (INIS)

    Abe, Kiyoharu.

    1995-05-01

    This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

  2. ACR-1000 design provisions for severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    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

  3. SARNET: Severe accident research network of excellence

    International Nuclear Information System (INIS)

    Albiol, T.; Van Dorsselaere, J. P.; Chaumont, B.; Haste, T.; Journeau, Ch.; Meyer, L.; Sehgal, Bal Raj; Schwinges, Bernd; Beraha, D.; Annunziato, A.; Zeyen, R.

    2010-01-01

    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

  4. Implementation of severe accident management measures - Summary and conclusions

    International Nuclear Information System (INIS)

    2002-01-01

    The objectives of the meeting were: 1) to exchange information on activities in the area of SAM implementation and on the rationale for such actions, 2) to monitor progress made, 3) to identify cases of agreement or disagreement, 4) to discuss future orientations of work, 5) to make recommendations to the CSNI. Session summaries prepared by the Chairpersons and discussed by the whole writing group are given in Annex. During the first session, 'SAM Programmes Implementation', papers from one regulator and several utilities and national research institutes were presented to outline the status of implementation of SAM programmes in countries like Switzerland, Russia, Spain, Finland, Belgium and Korea. Also, the contribution of SAM to the safety of Japanese plants (in terms of core damage frequency) was quantified in a paper. One paper gave an overview on the situation regarding SAM implementation in Europe. The second session, 'SAM Approach', provided background and bases for Severe Accident Management in countries like Sweden, Japan, Germany and Switzerland, as well as for hardware features in advanced light water reactor designs, such as the European Pressurised Reactor (EPR), regarding Severe Accident Management. The third session, 'SAM Mitigation Measures', was about hardware measures, in particular those oriented towards hydrogen mitigation where fundamentally different approaches have been taken in Scandinavian countries, France, Germany and Korea. Three papers addressed specific contributions from research to provide a broader basis for the assumptions made in certain computer codes used for the assessment of plant risk arising from beyond-design accident sequences. The fourth session, 'Implementation of SAM Measures on VVER-1000 Reactors', was about the status of work on Severe Accident Management implementation in VVER reactors of existing design and in a new plant currently under construction. The overall picture is that Severe Accident Management has been

  5. Big Rock Point severe accident management strategies

    International Nuclear Information System (INIS)

    Brogan, B.A.; Gabor, J.R.

    1996-01-01

    December 1994, the Nuclear Energy Institute (NEI) issued guidance relative to the formal industry position on Severe Accident Management (SAM) approved by the NEI Strategic Issues Advisory Committee on November 4, 1994. This paper summarizes how Big Rock Point (BRP) has and continues to address SAM strategies. The historical accounting portion of this presentation includes a description of how the following projects identified and defined the current Big Rock Point SAM strategies: the 1981 Level 3 Probabilistic Risk Assessment performance; the development of the Plant Specific Technical Guidelines from which the symptom oriented Emergency Operating Procedures (EOPs) were developed; the Control Room Design Review; and, the recent completion of the Individual Plant Evaluation (IPE). In addition to the historical presentation deliberation, this paper the present activities that continue to stress SAM strategies

  6. Aerosol transport in severe reactor accidents

    International Nuclear Information System (INIS)

    Fynbo, P.; Haeggblom, H.; Jokiniemi, J.

    1990-01-01

    Aerosol behaviour in the reactor containment was studied in the case of severe reactor accidents. The study was performed in a Nordic group during the years 1985 to 1988. Computer codes with different aerosol models were used for calculation of fission product transport and the results are compared. Experimental results from LACE, DEMONA and Marviken-V are compared with the calculations. The theory of aerosol nucleation and its influence on the fission product transport is discussed. The behaviour of hygroscopic aerosols is studied. The pool scrubbing models in the codes SPARC and SUPRA are reviewed and some knowledge in this field is assessed on the background of an international rewiew. (author) 60 refs

  7. SARNET: Severe accident research network of excellence

    International Nuclear Information System (INIS)

    Albiol, Thierry; Haste, Tim; Dorsselaere, Jean-Pierre van

    2007-01-01

    51 organizations network in SARNET (Severe Accident Research NETwork of Excellence) their capacities of research in order to resolve the most important remaining uncertainties for enhancing, in regard of Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). This project, co-funded by the European Commission (EC), 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 exists between the different R and D national programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the actors involved in SA research in Europe (plus Canada). To reach these objectives, all the organizations networked in SARNET contribute to a so-called Joint Programme of Activities (JPA), which consists in: Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; Harmonizing and re-orienting the research programmes; Jointly analysing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; Developing 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; Developing Scientific Databases, in which all the results of research programmes are stored in a common format (DATANET); Developing a common methodology for Probabilistic Safety Assessment (PSA) of NPPs; Developing courses and writing a text book on SA for students and researchers; Promoting personnel mobility between various European organizations. After the first period (2004-2008), co-funded by the EC, the network will progressively evolve toward self-sustainability. The bases for such an evolution, still under discussion

  8. Evaluation of severe accident environmental conditions taking accident management strategy into account for equipment survivability assessments

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Jeong, Ji Hwan; Na, Man Gyun; Kim, Soong Pyung

    2003-01-01

    This paper presents a methodology utilizing accident management strategy in order to determine accident environmental conditions in equipment survivability assessments. In case that there is well-established accident management strategy for specific nuclear power plant, an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for accident management strategy or action appropriately taken. For this work, three different tools are introduced; Probabilistic Safety Assessment (PSA) outcomes, major accident management strategy actions, and Accident Environmental Stages (AESs). In order to quantitatively investigate an applicability of accident management strategy to equipment survivability, the accident simulation for a most likely scenario in Korean Standard Nuclear Power Plants (KSNPs) is performed with MAAP4 code. The Accident Management Guidance (AMG) actions such as the Reactor Control System (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparing with those from previous normal accident simulation, especially focused on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages

  9. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  10. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    Fedorov, V.G.; Rogov, M.F.; Podshibyakin, A.K.; Fil, N.S.; Volkov, B.E.; Semishkin, V.P.

    1998-01-01

    Severe accident related requirements formulated in General Regulations for Nuclear Power Plant Safety (OPB-88), in Nuclear Safety Regulations for Nuclear Power Stations' Reactor Plants (PBYa RU AS-89) and in other NPP nuclear and radiation guides of the Russian Gosatomnadzor are analyzed. In accordance with these guides analyses of beyond design basis accidents should be performed in the reactor plant design. Categorization of beyond design basis accidents leading to severe accidents should be made on occurrence probability and severity of consequences. Engineered features and measures intended for severe accident management should be provided in reactor plant design. Requirements for severe accident analyses and for development of measures for severe accident management are determined. Design philosophy and proposed engineered measures for mitigation of severe accidents and decrease of radiation releases are demonstrated using examples of large, WWER-1000 (V-392), and medium size WWER-640 (V-407) reactor plant designs. Mitigation of severe accidents and decrease of radiation releases are supposed to be conducted on basis of consistent realization of the defense in depth concept relating to application of a system of barriers on the path of spreading of ionizing radiation and radioactive materials to the environment and a set of engineered measures protecting these barriers and retaining their effectiveness. Status of fulfilled by OKB Gidropress and other Russian organizations experimental and analytical investigations of severe accident phenomena supporting design decisions and severe accident management procedures is described. Status of the works on retention of core melt inside the WWER-640 reactor vessel is also characterized

  11. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Borondo, L.; Serrano, C.; Fiol, M.J.; Sanchez, A.

    2000-01-01

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  12. Probabilistic Assessment of Severe Accident Consequence in West Bangka

    Science.gov (United States)

    Sunarko; Su'ud, Zaki

    2017-07-01

    Probabilistic dose assessment for severe accident condition is performed for West Bangka area. Source-term from WASH-1400 reactor analysis is used as a conservative release scenario for 1000 MWe PWR. Seven groups of isotopes are used in the simulation based on core inventory and release fraction. Population distribution for Muntok district and the area within a 100 km radius is obtained from 2014 data. Meteorological data is provided through cyclic sampling from a database containing two-year site-specific hourly records in 2014-2015 periods. PC-COSYMA segmented plume dispersion code is used to investigate the assumed the consequence of the accident scenario. The result indicates that early or deterministic effect is important for areas close the release point while long-term or stochastic effect is related to population distribution and covers area of up to 100 km from the release point. The mean annual expected values for early mortality and late mortality for the population within 100 km radius from Muntok site are 2.38×10-4 yr -1 and 1.33×10-3 yr -1 respectively.

  13. Developing a knowledge base for the management of severe accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.; Jenkins, J.P.

    1986-01-01

    Prior to the accident at Three Mile Island, little attention was given to the development of procedures for the management of severe accidents, that is, accidents in which the reactor core is damaged. Since TMI, however, significant effort has been devoted to developing strategies for severe accident management. At the same time, the potential application of artificial intelligence techniques, particularly expert systems, to complex decision-making tasks such as accident diagnosis and response has received considerable attention. The need to develop strategies for accident management suggests that a computerized knowledge base such as used by an expert system could be developed to collect and organize knowledge for severe accident management. This paper suggests a general method which could be used to develop such a knowledge base, and how it could be used to enhance accident management capabilities

  14. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM 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 shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. 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. 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. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

  15. Severe accident sequences simulated at the Grand Gulf Nuclear Station

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1999-01-01

    Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

  16. Tchernobyl: a severe accident and its image

    International Nuclear Information System (INIS)

    Strazzulla, J.

    1996-01-01

    This paper gives a strong criticism about the false informations that were disseminated by the mass media immediately after the Tchernobyl accident. This accident is taken as an example to illustrate a common attitude in journalistic comments of geopolitical events. (J.S.). 1 photo

  17. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  18. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  19. Centrifugal Filtration System for Severe Accident Source Term Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Shu Chang; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this paper is to present the conceptual design of a filtration system that can be used to process airborne severe accident source term. Reactor containment may lose its structural integrity due to over-pressurization during a severe accident. This can lead to uncontrolled radioactive releases to the environment. For preventing the dispersion of these uncontrolled radioactive releases to the environment, several ways to capture or mitigate these radioactive source term releases are under investigation at KAIST. Such technologies are based on concepts like a vortex-like air curtain, a chemical spray, and a suction arm. Treatment of the radioactive material captured by these systems would be required, before releasing to environment. For current filtration systems in the nuclear industry, IAEA lists sand, multi-venturi scrubber, high efficiency particulate arresting (HEPA), charcoal and combinations of the above in NS-G-1-10, 4.143. Most if not all of the requirements of the scenario for applying this technology near the containment of an NPP site and the environmental constraints were analyzed for use in the design of the centrifuge filtration system.

  20. Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

    International Nuclear Information System (INIS)

    Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

    2015-01-01

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

  1. On severe accident hydrogen behaviour in Loviisa

    International Nuclear Information System (INIS)

    Okkonen, T.

    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.)

  2. 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.).

  3. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  4. Design and Development of a Severe Accident Training System

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Park, Sun Hee; Kim, Dong Ha

    2005-01-01

    The nuclear plants' severe accidents have two big characteristics. One is that they are very rare accidents, and the other is that they bring extreme conditions such as the high pressure and temperature in their process. It is, therefore, very hard to get the severe accident data, without inquiring that the data should be real or experimental. In fact, most of severe accident analyses rely on the simulation codes where almost all severe accident knowledge is contained. These codes are, however, programmed by the Fortran language, so that their output are typical text files which are very complicated. To avoid this kind of difficulty in understanding the code output data, several kinds of graphic user interface (GUI) programs could be developed. In this paper, we will introduce a GUI system for severe accident management and training, partly developed and partly in design stage

  5. 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)

  6. A study on the development of framework and supporting tools for severe accident management

    International Nuclear Information System (INIS)

    Chang, Hyun Sop

    1996-02-01

    Through the extensive research on severe accidents, knowledge on severe accident phenomenology has constantly increased. Based upon such advance, probabilistic risk studies have been performed for some domestic plants to identify plant-specific vulnerabilities to severe accidents. Severe accident management is a program devised to cover such vulnerabilities, and leads to possible resolution of severe accident issues. This study aims at establishing severe accident management framework for domestic nuclear power plants where severe accident management program is not yet established. Emphasis is given to in-vessel and ex-vessel accident management strategies and instrumentation availability for severe accident management. Among the various strategies investigated, primary system depressurization is found to be the most effective means to prevent high pressure core melt scenarios. During low pressure core melt sequences, cooling of in-vessel molten corium through reactor cavity flooding is found to be effective. To prevent containment failure, containment filtered venting is found to be an effective measure to cope with long-term and gradual overpressurization, together with appropriate hydrogen control measure. Investigation of the availability of Yonggwang 3 and 4 instruments shows that most of instruments essential to severe accident management lose their desired functions during the early phase of severe accident progression, primarily due to the environmental condition exceeded ranges of instruments. To prevent instrument failure, a wider range of instruments are recommended to be used for some severe accident management strategies such as reactor cavity flooding. Severe accidents are generally known to accompany a number of complex phenomena and, therefore, it is very beneficial when severe accident management personnel is aided by appropriately designed supporting systems. In this study, a support system for severe accident management personnel is developed

  7. Scenarios catalog for the graphical console for analysis of severe accidents visualization of OEs, NAEs and calculation of source term of the NPP-LV; Catalogo de escenarios para la consola grafica para analisis de accidentes severos visualizacion de OEs, NAEs y calculo del termino fuente de la CLV

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Mendoza R, M. E.; Tijerina S, F.; Garcia C, T., E-mail: manuel.mendoza@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    A nuclear power plant is operated at all times within the design criteria of structures, systems and components, and according to the operation technical specifications. For different areas of work of a nuclear power plant is necessary to carry out practices in which is useful to have the prediction of the thermo-hydraulic and radiological progression of scenarios that imply exceeding that design bases, even reaching the damage of the fuel in different degree. During the exercises and drills of the External Plan of Emergency Response, the projection of doses is done to exercise the different tasks of the plan. To make the projection of doses is required to have the radiological source term of the scenario on which is practiced. Because of this, was identified the convenience of having a catalog of scenarios for which the radiological source term was calculated. In 2004, a first version of the catalog was produced for a power of 2027 MW, and in 2011 the catalog was updated for extended power conditions, 2317 MW. Both versions were made using the severe accident simulator MAAP-3B. That catalog consists of a form and an optical storage device. The form contains tables and figures in which the characteristics of the scenario to be practiced are searched and the electronic files of the corresponding radiological source term are located in the storage device. Due to the recent development of the graphical console for analysis of severe accidents, visualization of OEs, NAEs and calculation of the source term for the nuclear power plant of Laguna Verde (NPP-LV) CoGrAAS, the catalog printed was replaced by an electronic catalog for the CoGrAAS. The new catalog retains the philosophy of the previous catalog, constituted by a wide collection of scenarios that involve different circumstances and phenomena, that can be used to practice different tasks during training exercises or simulacrums, and combined with the following advantages: the scenario selection is made from an

  8. Dose estimation and evaluation of protector measures for a power plant's accidents scenario, using geographical information system

    International Nuclear Information System (INIS)

    Costa, E.M.; Biagio, R.M.S.; Alves, R.N.

    1999-01-01

    Since the initial phase of a project of a nuclear plant several environmental studies are carried out, and a considerable amount of relevant information is generated. Therefore, there is an increasing need of an integrated analysis of this information in order to better evaluate the potential impact associated to hypothetical accident scenarios of such plants. This paper presents a case-study, in which a hypothetical accident scenario is analysed taking into account the environmental and populational information of the Brazilian nuclear power plants region by using a geographical information system. Important areas for planning of protective measures are identified to provide a basis for further analysis. (author)

  9. 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.

  10. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MELCOR 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. (authors)

  11. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

  13. 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.

  14. Phenomenology of severe accidents in BWR type reactors. First part

    International Nuclear Information System (INIS)

    Sandoval V, S.

    2003-01-01

    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)

  15. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

  16. Application of the accident management information needs methodology to a severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

    1989-11-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

  17. Strategies for the prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Ader, C.; Heusener, G.; Snell, V.G.

    1999-01-01

    The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

  18. A severe accident analysis for the system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  19. SEVERE ACCIDENT ISSUES RAISED BY THE FUKUSHIMA ACCIDENT AND IMPROVEMENTS SUGGESTED

    OpenAIRE

    SONG, JIN HO; KIM, TAE WOON

    2014-01-01

    This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accide...

  20. The management of severe accidents in modern pressure tube reactors

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Blahnik, C.; Snell, V.G.; Duffey, R.B.

    2007-01-01

    Advanced new reactor designs resist severe accidents through a balance between prevention and mitigation. This balance is achieved by designing to ensure that such accidents are very rare; and by limiting core damage progression and releases from the plant in the event of such rare accidents. These design objectives are supported by a suitable combination of probabilistic safety analysis, engineering judgment and experimental and analytical study. This paper describes the approach used for the Advanced CANDU Reactor TM -1000 (ACR-1000) design, which includes provisions to both prevent and mitigate severe accidents. The paper describes the use of PSA as a 'design assist' tool; the analysis of core damage progression pathways; the definition of the core damage states; the capability of the mitigating systems to stop and control severe accident events; and the severe accident management opportunities for consequence reduction. (author)

  1. Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

    The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

  2. 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)

  3. 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)

    1998-12-31

    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)

  4. Numerical Study of Severe Accidents on Containment Venting Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong [FNC Technology Co., Yongin (Korea, Republic of); Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  5. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek

    2014-01-01

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  6. Use of probabilistic safety analyses in severe accident management

    International Nuclear Information System (INIS)

    Neogy, P.; Lehner, J.

    1991-01-01

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs

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

    International Nuclear Information System (INIS)

    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 interface module. Main functions of

  8. 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

  9. ARAMIS project: A comprehensive methodology for the identification of reference accident scenarios in process industries

    International Nuclear Information System (INIS)

    Delvosalle, Christian; Fievez, Cecile; Pipart, Aurore; Debray, Bruno

    2006-01-01

    In the frame of the Accidental Risk Assessment Methodology for Industries (ARAMIS) project, this paper aims at presenting the work carried out in the part of the project devoted to the definition of accident scenarios. This topic is a key-point in risk assessment and serves as basis for the whole risk quantification. The first result of the work is the building of a methodology for the identification of major accident hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. The term 'major accidents' must be understood as the worst accidents likely to occur on the equipment, assuming that no safety systems are installed. A second methodology, called methodology for the identification of reference accident scenarios (MIRAS) takes into account the influence of safety systems on both the frequencies and possible consequences of accidents. This methodology leads to identify more realistic accident scenarios. The reference accident scenarios are chosen with the help of a tool called 'risk matrix', crossing the frequency and the consequences of accidents. This paper presents both methodologies and an application on an ethylene oxide storage

  10. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  11. 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...... principle of sustainable transit and advance the vision “every accident is one too many”. Methods Bus accident data were retrieved from the national accident database for the period 2002–2011. A generalized ordered logit model allows analyzing bus accident severity and a logistic regression enables...... examining occurrence of injury to bus passengers. Results Bus accident severity is positively related to (i) the involvement of vulnerable road users, (ii) high speed limits, (iii) night hours, (iv) elderly drivers of the third party involved, and (v) bus drivers and other drivers crossing in yellow or red...

  12. Severe accidents and nuclear containment integrity (SANCY). SANCY summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I. [VTT Processes, Espoo (Finland)

    2004-07-01

    SANCY project investigates physical phenomena related to severe nuclear accidents with importance to Finnish nuclear power plants. Currently the major topics are the ex-vessel coolability issues, long-term severe accident management and containment leak tightness and adoption and development of new calculation tools considering also the needs of the future Olkiluoto 3 plant. SANCY employs both experimental and analytical methods. (orig.)

  13. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  14. An overview of selected severe accident research and applications

    International Nuclear Information System (INIS)

    Hammersley, R.J.; Henry, R.E.

    2004-01-01

    Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

  15. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

    2013-12-01

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

  16. High-speed radiation dose calculations for severe accidents using INDOS

    International Nuclear Information System (INIS)

    Davidson, G.R.; Godin-Jacqmin, L.J.; Raines, J.C.

    1992-01-01

    The computer code INDOS (in-plant dose) has been developed for the high-speed calculation of in-plant radiation dose rates and doses during and/or due to a severe accident at a nuclear power plant. This paper describes the current capabilities of the code and presents the results of calculations for several severe-accident scenarios. The INDOS code can be run either as a module of MAAP, a code widely used in the nuclear industry for simulating the response of a light water reactor system during severe accidents, or as a stand-alone code using output from an alternative companion code. INDOS calculates gamma dose rates and doses in major plant compartments caused by airborne and deposited fission products released during an accident. The fission product concentrations are determined by the companion code

  17. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 2. Accident scenario based on thermodynamic model

    International Nuclear Information System (INIS)

    Maruyama, Shigenao

    2012-01-01

    An accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 2 is analyzed from the data open to the public. Phase equilibrium process model was introduced that the vapor and water are at saturation point in the vessels. Proposed accident scenario agrees very well with the data of the plant parameters obtained just after the accident. The estimation describes that the rupture time of the reactor pressure vessel (RPV) was at 22:50 14/3/2011. The estimation shows that the rupture time of the pressure containment vessel (RCP) was at 7:40 15/3/2011. These estimations are different from the ones by TEPCO, however; many measured evidences show good accordance with the present scenario. (author)

  18. Design study on dose evaluation method for employees at severe accident

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio; Kudo, Seiichi; Nishimura, Kazuya

    2001-01-01

    When we assume a severe accident in a nuclear power plant, it is required for rescue activity in the plant, accident management, repair work of failed parts and evaluation of employees to obtain radiation dose rate distribution or map in the plant and estimated dose value for the above works. However it might be difficult to obtain them accurately along the progress of the accident, because radiation monitors are not always installed in the areas where the accident management is planned or the repair work is thought for safety-related equipments. In this work, we analyzed diffusion of radioactive materials in case of a severe accident in a pressurized water reactor plant, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system by modeling a specific part of components and buildings in the plant from this design study on dose evaluation method for employees at severe accident, and then evaluated its availability. As a result, we obtained the followings: (1) A new dose evaluation method was established to predict the radiation dose rate in any point in the plant during a severe accident scenario. (2) This evaluation of total dose including moving route and time for the accident management and the repair work is useful for estimating radiation dose limit for these actions of the employees. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  19. Design study on dose evaluation method for employees at severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio [Institute of Nuclear Safety Systems Inc., Mihama, Fukui (Japan); Kudo, Seiichi [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)

    2001-09-01

    When we assume a severe accident in a nuclear power plant, it is required for rescue activity in the plant, accident management, repair work of failed parts and evaluation of employees to obtain radiation dose rate distribution or map in the plant and estimated dose value for the above works. However it might be difficult to obtain them accurately along the progress of the accident, because radiation monitors are not always installed in the areas where the accident management is planned or the repair work is thought for safety-related equipments. In this work, we analyzed diffusion of radioactive materials in case of a severe accident in a pressurized water reactor plant, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system by modeling a specific part of components and buildings in the plant from this design study on dose evaluation method for employees at severe accident, and then evaluated its availability. As a result, we obtained the followings: (1) A new dose evaluation method was established to predict the radiation dose rate in any point in the plant during a severe accident scenario. (2) This evaluation of total dose including moving route and time for the accident management and the repair work is useful for estimating radiation dose limit for these actions of the employees. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  20. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-07-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The accident sequence chosen for the calculating was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D was chosen with a wind speed of 5 m/s, and as extreme weather, Pasquill F with a wind speed of 2 m/s. 23 tabs., 37 ills., 20 refs. (author)

  1. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-01-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Ringhals reactor No 3/4. The accident sequence chosen for the calcualtions was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. A decontamination factor of 500 is used to account for the scrubber effect. Meteorological data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D, was chosen with a wind speed of 10 m/s, and as extreme weather, Pasquill E, with a wind speed of 2 m/s. 19 refs. (author)

  2. Aspects of severe accidents in transmutation systems

    International Nuclear Information System (INIS)

    Wider, H.U.; Karlson, J.; Jones, A.V.

    2001-01-01

    The different types of transmutation systems under investigation include accelerator driven (ADS) and critical systems. To switch off an accelerator in case of an accident initiation is quite important for all accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi has considerable advantages in this regard compared to gas cooling. Moreover it allows passive ex-vessel cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat capacity. (author)

  3. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

    The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

  4. Thermal-hydraulic uncertainties affecting severe accident progression

    International Nuclear Information System (INIS)

    Haskin, F.E.; Behr, V.L.

    1984-01-01

    To provide the proper technical bases for decisions regarding severe accidents, the US Nuclear Regulatory Commission (NRC) is sponsoring the following activities: (a) a variety of severe accident research programs, combined under the Severe Accident Research Plan; (b) nationwide task forces on containment loading, containment response, and fission product source terms; (c) a review by the American Physical Society of state-of-the-art methods for calculating radiological source terms; and (d) technical exchange meetings with the Industry Degraded Core (IDCOR) program. One of the means for integrating this developing array of technical information is the Severe Accident Risk Reduction Program (SARRP). One of the current SARRP objectives is to utilize insights gained from the activities listed above to characterize the relative likelihoods of competing containment failure modes for core-melt accidents

  5. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe accident research results to enhance the capability of plant operating personnel to effectively manage severe accidents. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed as part of the research program designed to resolve this issue. The methodology identifies the information needs of the plant personnel during a wide range of accident conditions, the existing plant measurements capable of supplying these information needs and what, if any minor additions to instrument and display systems would enhance the capability to manage accidents, known limitations on the capability of these measurements to function properly under the conditions that will be present during a wide range of severe accidents, and areas in which the information systems could mislead plant personnel. This paper presents an application of this methodology to a severe accident sequence to demonstrate its use in identifying the information which is available for management of the event. The methodology has been applied to a severe accident sequence in a Pressurized Water Reactor with a large dry containment. An examination of the capability of the existing measurements was then performed to determine whether the information needs can be supplied

  6. 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.

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

    International Nuclear Information System (INIS)

    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

  8. Addressing severe accidents in the CANDU 9 design

    International Nuclear Information System (INIS)

    Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

    1998-01-01

    CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

  9. Development on quantitative safety analysis method of accident scenario. The automatic scenario generator development for event sequence construction of accident

    International Nuclear Information System (INIS)

    Kojima, Shigeo; Onoue, Akira; Kawai, Katsunori

    1998-01-01

    This study intends to develop a more sophisticated tool that will advance the current event tree method used in all PSA, and to focus on non-catastrophic events, specifically a non-core melt sequence scenario not included in an ordinary PSA. In the non-catastrophic event PSA, it is necessary to consider various end states and failure combinations for the purpose of multiple scenario construction. Therefore it is anticipated that an analysis work should be reduced and automated method and tool is required. A scenario generator that can automatically handle scenario construction logic and generate the enormous size of sequences logically identified by state-of-the-art methodology was developed. To fulfill the scenario generation as a technical tool, a simulation model associated with AI technique and graphical interface, was introduced. The AI simulation model in this study was verified for the feasibility of its capability to evaluate actual systems. In this feasibility study, a spurious SI signal was selected to test the model's applicability. As a result, the basic capability of the scenario generator could be demonstrated and important scenarios were generated. The human interface with a system and its operation, as well as time dependent factors and their quantification in scenario modeling, was added utilizing human scenario generator concept. Then the feasibility of an improved scenario generator was tested for actual use. Automatic scenario generation with a certain level of credibility, was achieved by this study. (author)

  10. Using MARS to assist in managing a severe accident

    International Nuclear Information System (INIS)

    Raines, J.C.; Hammersley, R.J.; Henry, R.E.

    2004-01-01

    During an accident, information about the current and possible future states of the plant provides guidance for accident managers in evaluating which actions should be taken. However, depending upon the nature of the accident and the stress levels imposed on the plant staff responding to the accident the current and future plant assessments may be very difficult or nearly impossible to perform without supplemental training and/or appropriate tools. The MAAP Accident Response System (MARS) has been developed as a calculational aid to assist the responsible accident management individuals. Specifically MARS provides additional insights on the current and possible future states of the plant during an accident including the influence of operator actions. In addition to serving as a calculational aid, the MARS software can be an effective means for providing supplemental training. The MARS software uses engineering calculations to perform an integral assessment of the plant status including a consistency assessment of the available instrumentation. In addition, it uses the Modular Accident Analysis Program (MAAP) to provide near term predictions of the plant response if corrective actions are taken. This paper will discuss the types of information that are beneficial to the accident manager and how MARS addresses each. The MARS calculational functions include: instrumentation, validation and simulation, projected operator response based on the EOPs, as well as estimated timing and magnitude of in-plant and off-site radiation dose releases. Each of these items is influential in the management of a severe accident. (author)

  11. Airborne concentrations of radioactive materials in severe accidents

    International Nuclear Information System (INIS)

    Ross, D.F. Jr.; Denning, R.S.

    1989-01-01

    Radioactive materials would be released to the containment building of a commercial nuclear reactor during each of the stages of a severe accident. Results of analyses of two accident sequences are used to illustrate the magnitudes of these sources of radioactive materials, the resulting airborne mass concentrations, the characteristics of the airborne aerosols, the potential for vapor forms of radioactive materials, the effectiveness of engineered safety features in reducing airborne concentrations, and the release of radioactive materials to the environment. Ability to predict transport and deposition of radioactive materials is important to assessing the performance of containment safety features in severe accidents and in the development of accident management procedures to reduce the consequences of severe accidents

  12. BWR severe accident sequence analyses at ORNL - some lessons learned

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

    Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product

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

    International Nuclear Information System (INIS)

    2001-01-01

    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.)

  14. 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.)

  15. 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.)

  16. Strategy-oriented display concept to assist severe accident management

    International Nuclear Information System (INIS)

    Jeong, Kwangsub; Ha, Jaejoo

    2000-01-01

    The Critical Function Monitoring System (CFMS) is a typical Safety Parameter Display System (SPDS) to assist the operation of Korean Standard Nuclear Power Plants during normal and emergency operation, and SPDS for severe accident is being developed in Korea. When the existing CFMS is used under a severe accident situation, some problems are expected from: (1) different design basis, i.e. prevention of core melt vs. protection of radiation release to environment, (2) different parameters for decision-making, and (3) different domain and depth of information to restore the plant. To resolve the above problems, a concept, 'Strategy-Oriented Information Display' concept, for displaying information for severe accident management is developed in this paper. Whereas the existing SPDS structure is based on the critical safety function, the developed concept is based on the severe accident management strategy. The display for each strategy includes the plant parameters to check the status of plant and component with the logical or graphical views necessary for executing the strategy. As the application of the proposed concept, KAERI is developing a display system, the prototype severe accident SPDS, Severe Accident Management Display System (SAMDIS), to assist plant personnel for executing Korean Severe Accident Management Guidelines. CFMS is developed for a general display suitable to all situations with various displays. On the contrary, SAMDIS provides all the relevant information on one screen based on the proposed concept. The SAMDIS screen shows more extensive area than CFMS and thus plant personnel can recognize the overall plant status at a glance. This concept is quite effective when used with severe accident management guidelines because of the relatively macroscopic characteristics of a severe accident management strategy. (author)

  17. Economic consequences assessment for scenarios and actual accidents do the same methods apply

    International Nuclear Information System (INIS)

    Brenot, J.

    1991-01-01

    Methods for estimating the economic consequences of major technological accidents, and their corresponding computer codes, are briefly presented with emphasis on the basic choices. When applied to hypothetic scenarios, those methods give results that are of interest for risk managers with a decision aiding perspective. Simultaneously the various costs, and the procedures for their estimation are reviewed for some actual accidents (Three Mile Island, Chernobyl,..). These costs are used in a perspective of litigation and compensation. The comparison of the methods used and cost estimates obtained for scenarios and actual accidents shows the points of convergence and discrepancies that are discussed

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

    International Nuclear Information System (INIS)

    Bui, V.A.

    1998-01-01

    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

  19. Benchmarking severe accident computer codes for heavy water reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

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

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  1. The philosophy of severe accident management in the US

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1990-01-01

    The US NRC has put forth the initial steps in what is viewed as the resolution of the severe accident issue. Underlying this process is a fundamental philosophy that if followed will likely lead to an order of magnitude reduction in the risk of severe accidents. Thus far, this philosophy has proven cost effective through improved performance. This paper briefly examines this philosophy and the next step in closure of the severe accident issue, the IPE. An example of the authors experience with determinist. (author)

  2. Assessment of clad integrity of PHWR fuel pin following a postulated severe accident

    International Nuclear Information System (INIS)

    Dutta, B.K.; Kushwaha, H.S.; Venkat Raj, V.

    2000-01-01

    A mechanistic fuel performance analysis code FAIR has been developed. The code can analyse fuel pins with free standing as well as collapsible clad under normal, off-normal and accident conditions of reactors. The code FAIR is capable of analysing the effects of high burnup on fuel behaviour. The code incorporates finite element based thermo-mechanical module for computing transient temperature distribution and thermal-elastic-plastic stresses in the fuel pin. A number of high temperature thermo-physical and thermo-mechanical models also have been incorporated for analysing fuel pins subjected to severe accident scenario. The present paper describes salient features of code FAIR and assessment of clad integrity of PHWR fuel pins with different initial burnup subjected to severe accident scenario. (author)

  3. Estimation of cost per severe accident for improvement of accident protection and consequence mitigation strategies

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2013-01-01

    To assess the complex situations regarding the severe accidents such as what observed in Fukushima Accident, not only radiation protection aspects but also relevant aspects: health, environmental, economic and societal aspects; must be all included into the consequence assessment. In this study, the authors introduce the “cost per severe accident” as an index to analyze the consequences of severe accidents comprehensively. The cost per severe accident consists of various costs and consequences converted into monetary values. For the purpose of improvement of the accident protection and consequence mitigation strategies, the costs needed to introduce the protective actions, and health and psychological consequences are included in the present study. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 probabilistic safety assessment (PSA) codes. The accident sequences used in this analysis were taken from the results of level 2 seismic PSA of a virtual 1,100 MWe BWR-5. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR of Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the outputs are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated based on the open documents on the Fukushima Accident regarding the cost of protective actions and compensations for psychological harms. Finally, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. (author)

  4. Research on sever accident emergency simulation system for CPR1000

    International Nuclear Information System (INIS)

    Yang Zhifei; Liao Yehong; Liang Manchun; Li Ke; Yang Jie; Chen Yali

    2015-01-01

    The enhanced capability to nuclear power plant (NPP) severe accident management and emergency response depends heavily on exercises. Since the exercise scene is usually monotonous and not realistic, and conduct of exercise has a high cost, the effect of enhancing the capability is limited. Thus, the development of a Sever Accident Emergency Simulation System (SAESS) is necessary. SAESS is able to connect NPP simulator, and simulates the process of severe accident management, personnel evacuation, the dispersion of radioactive plume, and emergency response of emergency organizations. The system helps to design several of exercise scenes and optimize the disposal strategy in different severe accidents. In addition, the system reduces the cost of emergency exercise by computer simulation, benefits the research of exercise, increases the efficiency of exercise and enhances the emergency decision-making capability. This paper introduces the design and application of SAESS. (author)

  5. Geographic analysis of road accident severity index in Nigeria.

    Science.gov (United States)

    Iyanda, Ayodeji E

    2018-05-28

    Before 2030, deaths from road traffic accidents (RTAs) will surpass cerebrovascular disease, tuberculosis, and HIV/AIDS. Yet, there is little knowledge on the geographic distribution of RTA severity in Nigeria. Accident Severity Index is the proportion of deaths that result from a road accident. This study analysed the geographic pattern of RTA severity based on the data retrieved from Federal Road Safety Corps (FRSC). The study predicted a two-year data from a historic road accident data using exponential smoothing technique. To determine spatial autocorrelation, global and local indicators of spatial association were implemented in a geographic information system. Results show significant clusters of high RTA severity among states in the northeast and the northwest of Nigeria. Hence, the findings are discussed from two perspectives: Road traffic law compliance and poor emergency response. Conclusion, the severity of RTA is high in the northern states of Nigeria, hence, RTA remains a public health concern.

  6. Studies of severe accidents in light water reactors. Containment performance

    International Nuclear Information System (INIS)

    Hayns, M.R.; Phillips, D.W.; Young, R.L.D.

    1987-01-01

    The containment system of a LWR is an obvious component of the plant which performs an important safety function in preventing the release of fission products to the environment in the event of design basis accidents. With over 260 LWRs in service worldwide, and others still under construction, there is a considerable diversity of containment types and combinations of containment safeguards systems. All of these satisfy local regulatory requirements which are principally aimed at the design basis accidents, and these requirements naturally have a considerable uniformity. However, their design diversity becomes more relevant to the performance of the containment in severe accident conditions, and this aspect of containment performance is reviewed in this paper. The ability of the containment to mitigate severe accident consequences introduces the potential for accident management and recovery and this in turn points towards a range of new containment systems and concepts. PSA helps in judging these possibilities and in forming policies and procedures for accident management. It is perhaps in accident management that severe accident containment performance will be most beneficial in the future, and where additional effort in containment analysis will be focused

  7. An assessment the severe accident equipment survivability for the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Lee, B. C.; Moon, Y. T.; Park, J. W.; Kho, H. J.; Lee, S. W.

    1999-01-01

    One of the prominent design approaches to cope with the severe accident challenges in the Korean Next Generation Reactor is an assessment of equipment survivability in the severe accident environment at early design stage. In compliance with 10CFR50.34(f) and SECY-93-087, this work addresses that a reasonable level of assurance be provided to demonstrate that sufficient instrumentation and equipment will survive the consequences of a severe accident and will be available so that the operator may recover from and trend severe core damage sequences, including those scenarios which result in 100 percent oxidation of the active fuel cladding. An analytical and systematic approach was used to identify the equipment and instrumentation of safety-function and define severe accident environments including temperature, pressure, humidity, and radiation before and after the reactor vessel breach. As a result, it was concluded that with minor exceptions, existing design basis equipment qualification methods are sufficient to provide a reasonable level of assurance that this equipment will function during a severe accident. Furthermore, supplemental severe accident equipment and instrument procurement requirements were identified. (author)

  8. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsn, S.

    1987-03-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The assumption used for the calculations were a 0.06% release of iodine and cesium corresponding to a 0.1% release through the FILTRA plant at Barsebaeck. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather pasquill D was chosen with wind speed 5 m/s, and as extreme weather, Pasquill F with wind speed 2 m/s. 23 tabs., 36 ills., 21 refs. (author)

  9. Using MFM methodology to generate and define major accident scenarios for quantitative risk assessment studies

    DEFF Research Database (Denmark)

    Hua, Xinsheng; Wu, Zongzhi; Lind, Morten

    2017-01-01

    to calculate likelihood of each MAS. Combining the likelihood of each scenario with a qualitative risk matrix, each major accident scenario is thereby ranked for consideration for detailed consequence analysis. The methodology is successfully highlighted using part of BMA-process for production of hydrogen......Generating and defining Major Accident Scenarios (MAS) are commonly agreed as the key step for quantitative risk assessment (QRA). The aim of the study is to explore the feasibility of using Multilevel Flow Modeling (MFM) methodology to formulating MAS. Traditionally this is usually done based...

  10. Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

    International Nuclear Information System (INIS)

    Park, S.Y.; Ahn, K.I.

    2014-01-01

    Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

  11. The estimation economic impacts from severe accidents of a nuclear power plant

    International Nuclear Information System (INIS)

    Jeong, J. T.; Jeong, W. D.

    2001-01-01

    The severe accidents of a nuclear power plant may cause health effects in the exposed population and societal economic impacts or costs. Techniques to assess the consequences of an accident in terms of cost may be applied in studies on the design of plant safety features and in examining countermeasure options as part of emergency planning or in decision making after an accident. In this study, the costs resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of source term release parameters and meteorological data. Also, the costs were estimated for the different scenarios considering seasonal characteristics of Korea. The results can be used as essential inputs in costs/benefit analysis and in developing optimum risk reduction strategies

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

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

    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.)

  13. Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

    2005-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

  14. Leakage potential through mechanical penetrations in a severe accident environment

    International Nuclear Information System (INIS)

    Koenig, L.N.

    1986-01-01

    This paper reviews the findings of an ongoing program, Integrity of Containment Penetrations Under Severe Accident Loads. The program is concerned with the leakage modes as well as the magnitude of leakage through mechanical penetrations in a containment building subject to a severe accident. Seal and gasket tests are used to evaluate the effect of radiation aging, thermal aging, seal geometry, and seal squeeze on seals and gaskets subjected to a hypothesized severe accident. The effects on leakage of the structural response of equipment hatches, personnel airlocks, and drywell heads subjected to severe accident pressures are studied by experiments and analyses. The data gathered during this program will be used to develop methodologies for predicting leakage

  15. Severe accident management guidance for third Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Su Changsong

    2010-01-01

    The paper describes the background, document structure and the summaries of Severe Accident Management Guidance (SAMG) for Third Qinshan Nuclear Power Plant (TQNPP), and also introduces briefly some design features and their impacts on SAMG. (authors)

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

    African Journals Online (AJOL)

    ADOWIE PERE

    Spatial Analysis of Accident Spots Using Weighted Severity Index (WSI) and ... pedestrians avoiding the use of pedestrian bridges/aid even when they are available. ..... not minding an unforeseen obstruction, miscalculations and wrong break.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  18. 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).

  19. Comparative assessment of severe accident risks in the energy sector

    International Nuclear Information System (INIS)

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

    1997-01-01

    This paper addresses one of the major limitations of the current comparative studies of environmental and health impacts of energy systems, i.e. the treatment of severe accidents. The work covers technical aspects of severe accidents and thus primarily reflects an engineering perspective on the energy-related risk issues. The assessments concern full energy chains associated with fossil sources (coal, oil and gas), nuclear power and hydro power. A comprehensive severe accidents database has been established. Thanks to the variety of information sources used, it exhibits in comparison with other corresponding databases a far more extensive coverage of the energy-related accidents. For hypothetical nuclear accidents the probabilistic approach has been employed and extended to cover the economic consequences of power reactor accidents. Results of comparisons between the various energy chains are shown and discussed along with a number of current issues in comparative assessment of severe accidents. As opposed to the previous studies, the aim of the present work has been, to cover whenever possible, a relatively broad spectrum of damage categories of interest. (author) 5 figs., 1 tab., 18 refs

  20. Evaluation of strategies for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Tokarz, R.

    1989-01-01

    The NRC is planning to establish regulatory oversight on severe accident management capability in the US nuclear reactor industry. Accident management includes certain preparatory and recovery measures that can be taken by the plant operating and technical personnel to prevent or mitigate the consequences of a severe accident. Following an initiating event, accident management strategies include measures to (1) prevent core damage, (2) arrest the core damage if it begins and retain the core inside the vessel, (3) maintain containment integrity if the vessel is breached, and (4) minimize offsite releases. Objectives of the NRC Severe Accident Management Program are to assure that technically sound strategies are identified and guidance to implement these strategies is provided to utilities. This paper will describe work performed to date by Pacific Northwest Laboratory (PNL) and Battelle Memorial Institute (BMI) relative to severe accident strategy evaluation, as well as work to be performed and expected results. Working with Brookhaven National Laboratory, PNL evaluated a series of NRC suggested accident management strategies. The evaluation of these strategies was divided between PNL and Brookhaven National Laboratory and a similar paper will be presented by Brookhaven regarding their strategy evaluation. This paper will stress the overall safety issues related to the research and emphasize the strategies that are applicable to major safety issues. The relationship of these research activities to other projects is discussed, as well as planning for future changes in the direction of work to be undertaken

  1. Nuclear power plant severe accident research plan. Revision 1

    International Nuclear Information System (INIS)

    Marino, G.P.

    1986-04-01

    Subsequent to the Three Mile Island Unit 2 accident, recommendations were made by a number of review committees to consider regulatory changes which would provide better protection of the public from severe accidents. Over the past six years a major research effort has been underway by the NRC to develop an improved understanding of severe accidents and to provide a technical basis to support regulatory decisions. The purpose of this report is to describe current plans for the completion and extension of this research in support of ongoing regulatory actions in this area

  2. Simulation of severe accident using March-3 computer code

    International Nuclear Information System (INIS)

    Fernandes, A.; Nakata, H.

    1991-01-01

    The severe accident sensitivity analysis utilizing the March-3 approximate modelization options has been performed. The reference results against which the present results have been compared were obtained from the best published results for the most representative accident sequences: TMLU, S sub(2)DC sub(r) and S sub(2)DCF sub(r) for the Zion-1 reactor. The results of the present sensitivity analysis revealed the presence of very crude modelizations, in the March-3 program, to represent the critical phenomenologies involved in the severe accident sequences considered, even though large uncertainties must still be taken into account due primarily to the scarcity of the integral benchmark data. (author)

  3. Development of Krsko Severe Accident Management Guidance (SAMG)

    International Nuclear Information System (INIS)

    Cizel, F.

    1999-01-01

    In this lecture development of severe accident management guidances for Krsko NPP are described. Author deals with the history of severe accident management and implementation of issues (validation, review of E-plan and other aspects SAMG implementation guidance). Methods of Westinghouse owners group, of Combustion Engineering owners group, of Babcock and Wilcox owners group, of the BWR owners group, as well as application of US SAMG methodology in Europe and elsewhere are reviewed

  4. Simulator drills for the management of severe accidents

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1989-01-01

    The present state of deliberations on the simulation of severe accidents is presented and applied to a training philosophy. The special characteristics of 'severe' accidents are addressed and, falling under this category, the 'psychological structure of the man-machine-situation' is examined. The valid rules for drilling 'post-RESA-conduct' (RESA = fast reactor shut down) and the monitoring of safety goals are introduced. 2 figs., 1 tab

  5. 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.

  6. Severities of transportation accidents involving large packages

    International Nuclear Information System (INIS)

    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

  7. Severe accident modeling and offsite dose consequence evaluations for nuclear power plant emergency planning

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S.H.; Feng, T.S.; Huang, K.C. [National Tsing-Hua Univ., Hsinchu, Taiwan (China); Wang, J.R. [Inst. of Nuclear Energy Research, Longtan, Taiwan (China); Cheng, Y.H. [Industrial Tech. Res. Inst., Hsinchu, Taiwan (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing-Hua Univ., Hsinchu, Taiwan (China)

    2011-07-01

    We have investigated the roles of Firewater Addition System and Passive Flooder in ABWR severe accidents, such as LOCA and SBO. The results are apparent that Firewater System is vital in the highly unlikely situation where all AC are lost. Also in this paper, we present EPZDose, an effective and faster-than-real time code for offsite dose consequences predictions and evaluations. Illustrations with the release from our severe accident scenario show friendly and informative user's interface for supporting decision makings in nuclear emergency situations. (author)

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    -reporting. The problem of under-reporting is not unique for traffic accidents as severe under-reporting is a challenge in many other fields of incident reporting. In other incidents fields with intended or unintended harm, research has investigated the behavioural reasons for why people choose to report an incident......Under-reporting of traffic accidents is a well-discussed subject in traffic safety and it is well-known that the degree of under-reporting of traffic accidents is quite high in many countries. Nevertheless, very little literature has been made to investigate what causes the high degree of under...... on the service quality within the police none have looked at the service quality specific for the handling of traffic accidents.The objective of this Ph.D. thesis is to investigate the extent of under-reporting of traffic accidents in Denmark and trace the under-reporting systematically. As something new...

  9. ACCIDENT PHENOMENA OF RISK IMPORTANCE PROJECT - Continued RESEARCH CONCERNING SEVERE ACCIDENT PHENOMENA AND MANAGEMENT IN Sweden

    International Nuclear Information System (INIS)

    Rolandson, S.; Mueller, F.; Loevenhielm, G.

    1997-01-01

    Since 1988 all reactors in Sweden have mitigating measures, such as filtered vents, implemented. In parallel with the work of implementing these measures, a cooperation effort (RAMA projects) between the Swedish utilities and the Nuclear Power Inspectorate was performed to acquire sufficient knowledge about severe accident research work. The on-going project has the name Accident Phenomena of Risk Importance 3. In this paper, we will give background information about severe accident management in Sweden. In the Accident Phenomena of Risk Importance 3 project we will focus on the work concerning coolability of melted core in lower plenum which is the main focus of the In-vessel Coolability Task Group within the Accident Phenomena of Risk Importance 3 project. The Accident Phenomena of Risk Importance 3 project has joined on international consortium and the in-vessel cooling experiments are performed by Fauske and Associates, Inc. in Burr Ridge, Illinois, United States America, Sweden also intends to do one separate experiment with one instrument penetration we have in Swedish/Finnish BWR's. Other parts of the Accident Phenomena of Risk Importance 3 project, such as support to level 2 studies, the research at Royal Institute of Technology and participation in international programs, such as Cooperative Severe Accident Research Program, Advanced Containment Experiments and PHEBUS will be briefly described in the paper

  10. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  11. Structural evaluation of electrosleeved tubes under severe accident transients

    International Nuclear Information System (INIS)

    Majumdar, S.

    1999-01-01

    A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients

  12. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Moore, R.A.; Theofanous, T.G.

    1998-01-01

    The US Department of Energy (DOE) is sponsoring a programme in technology development aimed at resolving the technical issues in severe accident management strategies for advanced and evolutionary light water reactors (LWRs). The key objective of this effort is to achieve a robust defense-in-depth at the interface between prevention and mitigation of severe accidents. The approach taken towards this goal is based on the Risk Oriented Accident Analysis Methodology (ROAAM). Applications of ROAAM to the severe accident management strategy for the US AP600 advanced LWR have been effective both in enhancing the design and in achieving acceptance of the conclusions and base technology developed in the course of the work. This paper presents an overview of that effort and its key technical elements

  13. Instrumentation for the follow-up of severe accidents

    International Nuclear Information System (INIS)

    Munoz Sanchez, A.; Nino Perote, R.

    2000-01-01

    During severe accidents, it is foreseeable that the instrumentation installed in a plant is subjected to conditions which are more hostile than those for which the instrumentation was designed and qualified. Moreover, new, specific instrumentation is required to monitor variables which have not been considered until now, and to control systems which lessen the consequences of severe accidents. Both existing instrumentation used to monitor critical functions in design basis accident conditions and additional instrumentation which provides the information necessary to control and mitigate the consequences of severe accidents, have to be designed to withstand such conditions, especially in terms of measurements range, functional characteristics and qualification to withstand pressure and temperature loads resulting from steam explosion, hydrogen combustion/explosion and high levels of radiation over long periods of time. (Author)

  14. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-01-01

    This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research

  15. Revised Severe Accident Research Program plan, FY 1990--1992

    International Nuclear Information System (INIS)

    1989-08-01

    For the past 10 years, since the Three Mile Island accident, the NRC has sponsored an active research program on light-water-reactor severe accidents as part of a multi-faceted approach to reactor safety. This report describes the revised Severe Accident Research Program (SARP) and how the revisions are designed to provide confirmatory information and technical support to the NRC staff in implementing the staff's Integration Plan for Closure of Severe Accident Issues as described in SECY-88-147. The revised SARP addresses both the near-term research directed at providing a technical basis upon which decisions on important containment performance issues can be made and the long-term research needed to confirm and refine our understanding of severe accidents. In developing this plan, the staff recognized that the overall goal is to reduce the uncertainties in the source term sufficiently to enable the staff to make regulatory decisions on severe accident issues. However, the staff also recognized that for some issues it may not be practical to attempt to further reduce uncertainties, and some regulatory decisions or conclusions will have to be made with full awareness of existing uncertainties. 2 figs., 1 tab

  16. Validation of severe accident management guidance for the wolsong plants

    International Nuclear Information System (INIS)

    Park, S. Y.; Jin, Y. H.; Kim, S. D.; Song, Y. M.

    2006-01-01

    Full text: Full text: The severe accident management(SAM) guidance has been developed for the Wolsong nuclear power plants in Korea. The Wolsong plants are 700MWe CANDU-type reactors with heavy water as the primary coolant, natural uranium-fueled pressurized, horizontal tubes, surrounded by heavy water moderator inside a horizontal calandria vessel. The guidance includes six individual accident management strategies: (1) injection into primary heat transport system (2) injection into calandria vessel (3) injection into calandria vault (4) reduction of fission product release (5) control of reactor building condition (6) reduction of reactor building hydrogen. The paper provides the approaches to validate the SAM guidance. The validation includes the evaluation of:(l) effectiveness of accident management strategies, (2) performance of mitigation systems or components, (3) calculation aids, (4) strategy control diagram, and (5) interface with emergency operation procedure and with radiation emergency plan. Several severe accident sequences with high probability is selected from the plant specific level 2 probabilistic safety analysis results for the validation of SAM guidance. Afterward, thermal hydraulic and severe accident phenomenological analyses is performed using ISAAC(Integrated Severe Accident Analysis Code for CANDU Plant) computer program. Furthermore, the experiences obtained from a table-top-drill is also discussed

  17. Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

    International Nuclear Information System (INIS)

    Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

    2004-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

  18. Severity of electrical accidents in the construction industry in Spain.

    Science.gov (United States)

    Suárez-Cebador, Manuel; Rubio-Romero, Juan Carlos; López-Arquillos, Antonio

    2014-02-01

    This paper analyzes the severity of workplace accidents involving electricity in the Spanish construction sector comprising 2,776 accidents from 2003 to 2008. The investigation considered the impact of 13 variables, classified into 5 categories: Personal, Business, Temporal, Material, and Spatial. The findings showed that electrical accidents are almost five times more likely to have serious consequences than the average accident in the sector and it also showed how the variables of age, occupation, company size, length of service, preventive measures, time of day, days of absence, physical activity, material agent, type of injury, body part injured, accident location, and type of location are related to the severity of the electrical accidents under consideration. The present situation makes it clear that greater effort needs to be made in training, monitoring, and signage to guarantee a safe working environment in relation to electrical hazards. This research enables safety technicians, companies, and government officials to identify priorities and to design training strategies to minimize the serious consequences of electrical accidents for construction workers. Copyright © 2013 Elsevier Ltd and National Safety Council. All rights reserved.

  19. Recent Developments in Level 2 PSA and Severe Accident Management

    International Nuclear Information System (INIS)

    Ang, Ming Leang; Shepherd, Charles; Gauntt, Randall; Landgren, Vickie; Van Dorsselaere, Jean Pierre; Chaumont, Bernard; Raimond, Emmanuel; Magallon, Daniel; Prior, Robert; Mlady, Ondrej; Khatib-Rahbar, Mohsen; Lajtha, Gabor; Tinkler, Charles; Siu, Nathan

    2007-01-01

    In 1997, CSNI WGRISK produced a report on the state of the art in Level 2 PSA and severe accident management - NEA/CSNI/R(1997)11. Since then, there have been significant developments in that more Level 2 PSAs have been carried out worldwide for a variety of nuclear power plant designs including some that were not addressed in the original report. In addition, there is now a better understanding of the severe accident phenomena that can occur following core damage and the way that they should be modelled in the PSA. As requested by CSNI in December 2005, the objective of this study was to produce a report that updates the original report and gives an account of the developments that have taken place since 1997. The aim has been to capture the most significant new developments that have occurred rather than to provide a full update of the original report, most of which is still valid. This report is organised using the same structure as the original report as follows: Chapter 2: Summary on state of application, results and insights from recent Level 2 PSAs. Chapter 3: Discussion on key severe accident phenomena and modelling issues, identification of severe accident issues that should be treated in Level 2 PSAs for accident management applications, review of severe accident computer codes and the use of these codes in Level 2 PSAs. Chapter 4: Review of approaches and practices for accident management and SAM, evaluation of actions in Level 2 PSAs. Chapter 5: Review of available Level 2 PSA methodologies, including accident progression event tree / containment event tree development. Chapter 6: Aspects important to quantification, including the use of expert judgement and treatment of uncertainties. Chapter 7: Examples of the use of the results and insights from the Level 2 PSA in the context of an integrated (risk informed) decision making process

  20. 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

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

    International Nuclear Information System (INIS)

    Park, S. Y.; Song, Y. M.

    2015-01-01

    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

  2. Japanese FR Deployment Scenario Study after the Fukushima Accident

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Shiotani, Hiroki; Ohtaki, Akira; Mukaida, Kyoko; Abe, Tomoyuki

    2013-01-01

    Conclusion: • The results revealed a need for the implementation of reprocessing and development of FR from the view point of reducing waste, etc. in the medium to long term. • JAEA’s contribution to international cooperation and standardization focusing on the enhancement of safety and reduction of the radioactive waste burden will be increasingly important in current situation. • JAEA intend to continue to build and propose FR deployment scenarios and identify their characteristics

  3. 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.

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

    International Nuclear Information System (INIS)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

    2014-01-01

    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

  5. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  6. Extension of emergency operating procedures for severe accident management

    International Nuclear Information System (INIS)

    Chiang, S.C.

    1992-01-01

    To enhance the capability of reactor operators to cope with the hypothetical severe accident its the key issue for utilities. Taiwan Power Company has started the enhancement programs on extension of emergency operating procedures (EOPs). It includes the review of existing LOPs based on the conclusions and recommendations of probabilistic risk assessment studies to confirm the operator actions. Then the plant specific analysis for accident management strategy will be performed and the existing EOPs will be updated accordingly

  7. Analyzing the severity of accidents on the German Autobahn.

    Science.gov (United States)

    Manner, Hans; Wünsch-Ziegler, Laura

    2013-08-01

    We study the severity of accidents on the German Autobahn in the state of North Rhine-Westphalia using data for the years 2009 until 2011. We use a multinomial logit model to identify statistically relevant factors explaining the severity of the most severe injury, which is classified into the four classes fatal, severe injury, light injury and property damage. Furthermore, to account for unobserved heterogeneity we use a random parameter model. We study the effect of a number of factors including traffic information, road conditions, type of accidents, speed limits, presence of intelligent traffic control systems, age and gender of the driver and location of the accident. Our findings are in line with studies in different settings and indicate that accidents during daylight and at interchanges or construction sites are less severe in general. Accidents caused by the collision with roadside objects, involving pedestrians and motorcycles, or caused by bad sight conditions tend to be more severe. We discuss the measures of the 2011 German traffic safety programm in the light of our results. Copyright © 2013 Elsevier Ltd. All rights reserved.

  8. Overview of LWR severe accident research activities at the Karlsruhe Institute of Technology

    International Nuclear Information System (INIS)

    Miassoedov, Alexei; Albrecht, Giancarlo; Foit, Jerzy-Jan; Jordan, Thomas; Steinbrück, Martin; Stuckert, Juri; Tromm, Walter

    2012-01-01

    The research activities in the light water reactor (LWR) severe accidents domain at Karlsruhe Institute of Technology (KIT) are concentrated on the in- and ex-vessel core melt behavior. The overall objective is to investigate the core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity and to the containment, corium concrete interaction and corium coolability in the reactor cavity, and hydrogen behaviour in reactor systems. The results of the experiments contribute to a better understanding of the core melt sequences and thus improve safety of existing and, in the long-term, of future reactors by severe accident mitigation measures and by safety installations where required. This overview paper describes the experimental facilities used at KIT for severe accident research and gives an overview of the main directions and objectives of the R&D work. (author)

  9. Simulation of operator's actions during severe accident management

    International Nuclear Information System (INIS)

    Viktorov, A.

    2015-01-01

    Implementing accident management counter measures or actions to mitigate consequences of a severe accident is essential to reduce radiological risks to the public and environment. Station-specific severe accident management guidelines (SAMGs) have been developed and implemented at all Canadian nuclear power plants. Following the Fukushima Daiichi nuclear accident certain enhancements were introduced to the SAMG, namely consideration of multi-units accidents, events involving spent fuel pools, incorporation of capability offered by the portable emergency mitigating equipment, and so on. To evaluate the adequacy and usability of the SAMGs, CNSC staff initiated a number of activities including a desktop review of SAMG documentation, evaluation of SAMG implementation through exercises and interviews with station staff, and independent verification of SAMG action effectiveness. This paper focuses on the verification of SAMG actions through analytical simulations. The objectives of the work are two-folds: (a) to understand the effectiveness of SAMG-specified mitigation actions in addressing the safety challenges and (b) to check for potential negative effects of the action. Some sensitivity calculations were performed to help understanding of the impact from actions that rely on the partially effective equipment or limited material resources. The severe accident computer code MAAP4-CANDU is used as a tool in this verification. This paper will describe the methodology used in the verification of SAMG actions and some results obtained from simulations. (author)

  10. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    Silverman, Eugene B.

    2004-01-01

    SAMSON is a computational tool used by accident managers in the Technical Support Centers (TSC) and Emergency Operations Facilities (EOF) in the event of a nuclear power plant accident. SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. SAMSON uses expert systems, as well as neural networks trained with the back propagation learning algorithms to make predictions. Training on data from an accident analysis code (MAAP - Modular Accident Analysis Program) allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTRs), with breaks ranging from one tube to eight tubes, and loss of coolant accidents (LOCAs), with breaks ranging from 0.0014 square feet (1.30 cm 2 ) in size to breaks 3.0 square feet in size (2800 cm 2 ). (author)

  11. Fan Cooler Operation in Kori 1 for Mitigating Severe Accident

    International Nuclear Information System (INIS)

    Suh, Nam Duk; Park, Jae Hong

    2005-01-01

    The Korea Ministry of Science and Technology (MOST) issued the 'Policy on Severe Accident of Nuclear Power Plants' in August 2001. According to the policy it was required for the licensee to develop a plant specific severe accident management guideline (SAMG) and to implement it. Thus the utility has made an implementation plan to develop SAMGs for operating plants. The SAMG for Kori unit 1 was submitted to the government on January 2004. Since then, the government trusted KINS to review the submitted SAMG in view of its feasibility and effectiveness. The first principle of the developed SAMG is to use only the available facilities as it is without introducing any system change. Because Kori-1 has no mitigative facility against combustible gases during severe accident, it relies heavily both on spray and on fan cooler systems to control the containment condition. Thus one of the issues raised during the review is to know whether the fan coolers which are designed for DBA LOCA can be effective in mitigating the severe accident conditions. This paper presents an analysis result of fan cooler operation in controlling the containment condition during severe accident of Kori 1

  12. Analysis of Hydrogen Control Strategy Using Igniter during Severe Accident

    International Nuclear Information System (INIS)

    Lee, Sung Bok; Kim, Hyeong Taek; Lee, Keo Hyoung

    2008-01-01

    The Severe Accident Management Guidelines (SAMGs) for the operating pressurized water reactor (PWR) have been completed within 2006. Among the SAMG strategies, mitigation-07 is the most important strategy for managing a severe accident of a PWR in order to reduce containment hydrogen. The fastest way to reduce the containment hydrogen concentration is to intentionally ignite the hydrogen. For this strategy, igniters exist in Optimized Power Reactor 1000 (OPR 1000) to burn hydrogen for a severe accident. For using the igniters during a severe accident, the adverse effects such as the explosion of the hydrogen mixture should be considered for containment integrity. However, an applicable discrimination method to activate the igniters does not exist, so that the hydrogen control strategy using the igniters cannot be chosen during a severe accident. Thus, this study focused on suggesting an applicable discrimination method to carry out the strategy of using the igniters. In this study, the specific plant used for this analysis is Ulchin Unit 5 and 6, OPR 1000 plant, in Korea

  13. Design features of ACR in severe accident mitigation

    International Nuclear Information System (INIS)

    Shapiro, H.; Krishnan, V.S.; Santamaura, P.; Lekakh, B.; Blahnik, C.

    2007-01-01

    New reactor designs require the evaluation of design alternatives to reduce the radiological risk by preventing severe accidents or by limiting releases from the plant in the event of such accidents. The Advanced CANDU Reactor TM (ACR TM ) design has provisions to prevent and mitigate severe accidents. This paper describes key ACR design features for severe accident mitigation. It provides a high-level overview of the findings to date. Several design provisions have not yet been finalized or decided, but the designers are keenly aware of the SAM concepts and their requirements. The active heat sinks for 'vessels' (i.e., the fuel channels, the calandria vessel, the calandria end-shields and the calandria vault) are all amply capable of dissipating the severe accident heat loads. These heat sinks are designed to be operable under severe accident environmental conditions; however, their operability is yet to be confirmed by assessments. The active heat sinks for the various process vessels are 'backed up' by passive heat sinks (i.e., steaming plus water make-up from the RWS). The supply side of passive heat sinks is simple, rugged, and not vulnerable to failures of plant systems. The importance of the steam relief side is recognized, and the adequate relief capacity will be provided. The passive heat sinks will give the SAM more than 1 day (likely several days) to diagnose the accident and to establish the ultimate heat sinks. The spray system for containment pressure suppression is designed for high reliability and has ample capacity to ensure low containment leakage without external intervention, after which time alternative supply to the sprays can be brought on line manually. The sprays are backed up by the LACs which are assessed for operability following a severe accident. The strong ACR containment will provide a long time of completely passive protection for any severe accident at decay power. Its characteristics are not prone to catastrophic failures. The

  14. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  15. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

    2010-01-01

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

  16. The Tchernobyl enigma or: the human factors in severe accidents

    International Nuclear Information System (INIS)

    Llory, M.

    1988-01-01

    Using the analysis of many documents published after the Tchernobyl accident, we attempt to distinguish the main human factors aspects in severe accidents that come out, and the causes the most frequently quoted to ''explain'' it. But the Tchernobyl accident keeps its ''enigmatic'' feature, like any other accident. The need to make a deeper investigation concerning safety leads to look for various research paths that go beyond the usual normative positions, based on a too much mechanistic model of man. It is to the functioning of groups in work situations that we suggest to devote part of the research and thinking effort. We attempt to show briefly how two theories, the theory of ''groupthink'' and the theory of ''trade defensive ideologies'', can throw a light on the problem of human factors in nuclear power plants [fr

  17. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.; Manilia, E.; Pietrangeli, G.; Sennis, C.

    1985-01-01

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  18. Severe accident management: radiation dose control, Fukushima Daiichi and TMI-2 nuclear plant accidents

    International Nuclear Information System (INIS)

    Shaw, Roger

    2014-01-01

    This presentation presents valuable dose information related to the Fukushima Daiichi and Three Mile Island Unit 2 (TMI-2) Nuclear Plant accidents. Dose information is provided for what is well known for TMI-2, and what is available for Fukushima Daiichi. Particular emphasis is placed on the difference between the type of reactors involved, overarching plant damage issues, and radiation worker dose outcomes. For TMI-2, more in depth dose data is available for the accident and the subsequent recovery efforts. The comparisons demonstrate the need to understand the wide variation in potential dose management measures and outcomes for severe reactor accidents. (author)

  19. 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)

  20. A framework for assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems and actions to prevent or mitigate a severe accident. Together with risk management (changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-in-depth safety philosophy for severe accidents. A significant number of probabilistic safety assessments (PSA) have been completed which yield the principal plant vulnerabilities. For each sequence/threat and each combination of strategy there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainty in key phenomena, uncertainty in operator behavior, uncertainty in system availability and behavior, and uncertainty in available information (i.e., instrumentation). The objective of this project is to develop a methodology for assessing severe accident management strategies given the key uncertainties mentioned above. Based on Decision Trees and Influence Diagrams, the methodology is currently being applied to two case studies: cavity flooding in a PWR to prevent vessel penetration or failure, and drywell flooding in a BWR to prevent containment failure

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems and actions to prevent or mitigate a severe accident. Together with risk management (changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-in-depth safety philosophy for severe accidents. A significant number of probabilistic safety assessments (PSA) have been completed which yield the principal plant vulnerabilities. For each sequence/threat and each combination of strategy there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainty in key phenomena, uncertainty in operator behavior, uncertainty in system availability and behavior, and uncertainty in available information (i.e., instrumentation). The objective of this project is to develop a methodology for assessing severe accident management strategies given the key uncertainties mentioned above. Based on decision trees and influence diagrams, the methodology is currently being applied to two case studies: cavity flooding in a pressurized water reactor to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent containment failure

  2. Design consideration on severe accident for future LWR

    International Nuclear Information System (INIS)

    Omoto, A.

    1998-01-01

    Utilities' Severe Accident Management strategies, selected based on Individual Plant Examination, are in the process of implementation for each operating plant. Activities for the next generation LWR design are going on by Utilities, NSSS vendors and Research Institutes. The proposed new designs vary from evolutionary design to revolutionary design such as the supercritical LWR. Discussion on the consideration of Severe Accident in the design of next generation LWR is being held to establish the industry's self-regulatory document on containment design and its performance, which ABWR-IER (Improved Evolutionary Reactor) on the part of BWR and Evolutionary APWR and New PWR21 on the part of PWR are expected to comply. Conceptual design study for ABWR-IER will illustrate an example of design approach for the prevention and mitigation of Severe Accident and its impact on capital cost

  3. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  4. Plant specific severe accident management - the implementation phase

    International Nuclear Information System (INIS)

    Prior, R.

    1999-01-01

    Many plants are in the process of developing on-site guidance for technical staff to respond to a severe accident situation severe accident management guidance (SAMG). Once the guidance is developed, the SAMG must be implemented at the plant site, and this involves addressing a number of additional aspects. In this paper, approaches to this implementation phase are reviewed, including review and verification of plant specific SAMG, organizational aspects and integration with the emergency plan, training of SAMG users, validation and self-assessment and SAMG maintenance. Examples draw on experience from assisting numerous plants to implement symptom based severe accident management guidelines based on the Westinghouse Owners Group approach, in Westinghouse, non-Westinghouse and VVER plant types. It is hoped that it will be of use to those plant operators about to perform these activities.(author)

  5. 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

  6. Technique of research of severe accidents and substantiation of safety of nuclear systems

    International Nuclear Information System (INIS)

    Ivanov, E.A.; Tchenov, S.V.

    2001-01-01

    Work is devoted to development of possible ways of solution of the problems of nuclear safety substantiation. We believe that safety in severe accidents is one of significant factors, which restrict value of nuclear industry in future power production. In connection with it we can conclude followed items: -) Substantiation of safety in severe accidents in nuclear system should be built on a deterministic way of guaranteed exception of heavy consequences; -) It is easy that this aim can be achieved by modeling in functions of common type; -) Main purpose of this work is to show that it is possible to estimate physical allowed state of system in emergency and find of trajectory of heaviest scenarios by optimization procedure; and -) In this work we have developed new method and computer code purposed for study of accident conditions of water cooled un-managed nuclear systems such as cooling ponds of spent fuel, experimental facilities etc. (authors)

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

    International Nuclear Information System (INIS)

    Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

    1984-01-01

    The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure

  8. Severe accident considerations in Canadian nuclear power reactors

    International Nuclear Information System (INIS)

    Omar, A.M.; Measures, M.P.; Scott, C.K.; Lewis, M.J.

    1990-08-01

    This paper describes a current study on severe accidents being sponsored by the Atomic Energy Control Board (AECB) and provides background on other related Canadian work. Scoping calculations are performed in Phase I of the AECB study to establish the relative consequences of several permutations resulting from six postulated initiating events, nine containment states, and a selection of meteorological conditions and health effects mitigating criteria. In Phase II of the study, selected accidents sequences would be analyzed in detail using models suitable for the design features of the Canadian nuclear power reactors

  9. 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

  10. Formulating the Canadian regulatory position on severe accidents

    International Nuclear Information System (INIS)

    Viktorov, Alex

    2006-01-01

    In response to the increasing potential of new nuclear build in Canada, and as part of documentation harmonization effort, CNSC staff has initiated development of requirements for design of nuclear power plants. These requirements build both on the IAEA standards, most notably, NS-R-1, and the Canadian practices and experience. The three safety objectives, formulated by the IAEA, are adopted, and Safety Goals are proposed consistent with the international trend. This Canadian standard will require, for the first time, explicit consideration of severe accidents in design and safety assessments. Specific requirements are formulated for several plant systems that assure an effective fourth level of defence in depth. Available results from probabilistic safety assessments indicate that the risks posed by severe accidents are acceptably low. Nevertheless, such risks are not negligible. CNSC staff considers that severe accident management (SAM) represents the most practical way to achieve risk reduction with a moderate effort. Ultimately, SAM actions are aimed at bringing the reactor, and the plant in general, into a controlled and stable state. For the operating reactors, SAM provides an additional defense barrier against the consequences of those accidents that fall beyond the scope of events considered in the reactor design basis. The establishment of a SAM program ensures availability of the information, procedures, and resources necessary to take full advantage of existing plant capabilities to arrest core degradation, and prevent or mitigate large releases of radioactive material. To the extent practicable, a SAM program builds on the existing emergency operating procedures and makes use of the plant design capabilities. On this basis, the CNSC requested nuclear power reactor licensees to develop and implement SAM at all operating reactors. To be able to demonstrate compliance with requirements for plant design and severe accident management, it is necessary to

  11. Analytical measurements of fission products during a severe nuclear accident

    Science.gov (United States)

    Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

    2018-01-01

    The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  12. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    1987-01-01

    From 10 to 12 November 1986 some 80 delegates met under the auspices of the CEC working group on the safety of light-water reactors. The participants from EC Member States were joined by colleagues from Sweden, Finland and the USA and met to discuss the subject of severe accidents in LWRs. Although this seminar had been planned well before Chernobyl, the ''severe-accident-that-really-happened'' made its mark on the seminar. The four main seminar topics were: (i) high source-term accident sequences identified in PSAs, (ii) containment performance, (iii) mitigation of core melt consequences, (iv) severe accident management in LWRs. In addition to the final panel discussion there was also a separate panel discussion on lessons learned from the Chernobyl accident. These proceedings include the papers presented during the seminar and they are arranged following the seminar programme outline. The presentations and discussions of the two panels are not included in the proceedings. The general conclusions and directions following from these two panels were, however, considered in a seminar review paper which was published in the March 1987 issue of Nuclear Engineering International

  13. Analytical measurements of fission products during a severe nuclear accident

    Directory of Open Access Journals (Sweden)

    Doizi D.

    2018-01-01

    Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  14. Severe Accidents: French Regulatory Practice for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Colin, M.

    1997-01-01

    In the framework of a continuous and iterative process, the French Safety Authority asks the utility EDF to implement equipment and procedure modifications on the operating reactors, in order to cope with the most likely Severe Accident sequences. As a result of Probabilistic Safety Assessments published in 1990, important equipment and procedure modifications are being implemented on the French PWRs to improve the safety in shutdown states. The implementation of another set of modifications against some reactivity accident sequences is also in progress. More recently, the Safety Authority expressed specific Severe Accident requirements in terms of instrumentation, equipment qualification, high pressure core melt accidents and hydrogen risk prevention. In that respect, EDF was asked to implement hydrogen recombiners on its reactors. On the other hand, the French Safety authority is involved with its German counterpart in the assessment process of the European Pressurized Water Reactor Project. In consistency with the common recommendations of the Safety Authorities involved, Severe Accident provisions for this reactor are being taken into account at the design stage

  15. 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.

  16. Severe accidents and operator training - discussion of potential issues

    International Nuclear Information System (INIS)

    Vidard, Michel

    1997-01-01

    R and D programs developed throughout the world allowed significant progress in the understanding of physical phenomena and Severe Accident Management (SAM) programs started in many OECD countries. Basically, the common denominator to all these SAM programs was to provide utility operators with procedures or guidelines allowing to deal with complex situations not formally considered in the Design Basis, including accidents where a significant portion of the core had molten. These SAM procedures or guidelines complement the traditional accident management procedures (event, symptom or physical-state oriented) and should allow operators to deal with a reasonably bounding set of situations. Dealing with operator or crisis team training, it was recognized that training would be beneficial but that training programs were lagging, i.e. though training sessions were either organized or contemplated after implementation of SAM programs, they seemed to be somewhat different from more traditional training sessions on Accident Management. After some explanations on the differences between Design Basis Accidents (DBAs) and Beyond Design Basis Accidents (BDBAs), this paper underlines some potential difficulties for training operators and discuss problems to be addressed by organisms contemplating SAM training sessions consistent with similar activities for less complex events

  17. Development and assessment of ASTEC code for severe accident simulation

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.

    2005-01-01

    Full text of publication follows: The ASTEC integral code, jointly developed by IRSN and GRS since several years for evaluation of source term during a severe accident (SA) in a Light Water Reactor, will play a central role in the SARNET network of excellence of the 6. Framework Programme (FwP) of the European Commission which started in spring 2004. It should become the reference European SA integral code in the next years. The version V1.1, released in June 2004, allows to model most of the main physical phenomena (except steam explosion) near or at the state of the art. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time usually takes less than one day of real time to be simulated on a PC computer. Important efforts are being made on validation by covering more than 30 reference experiments, often International Standard Problems from OECD (CORA, LOFT, PACTEL, BETA, VANAM, ACE-RTF, Phebus.FPT1...). The code is also used for the detailed interpretation of all the integral Phebus.FP experiments. Eighteen European partners performed a first independent evaluation of the code capabilities in 2000-03 within the frame of the EVITA 5. FwP project on one hand by comparison to experiments and on another hand by benchmarking with MAAP4 and MELCOR integral codes on plant applications on PWR and VVER. Their main conclusions were the needs of improvement of code robustness (especially the 2 new modules CESAR and DIVA simulating respectively circuit thermal hydraulics and core degradation) and of post-processing tools. Some improvements have already been achieved in the latest version V 1.1 on these two aspects. A new module MEDICIS devoted to Molten Core Concrete Interaction (MCCI) is implemented in this version, with a tight coupling to the containment thermal hydraulics module CPA. The paper presents a detailed analysis of a TMLB sequence on a French 900 MWe PWR, from

  18. The technical requirements concerning severe accident management in nuclear power plants

    International Nuclear Information System (INIS)

    Okamoto, Koji; Sugiyama, Tomoyuki; Kamata, Shinya

    2014-01-01

    The Great East Japan Earthquake with a magnitude of 9.0 (The 2011 off the Pacific coast of Tohoku Earthquake) occurred on March 11, 2011, and the beyond design-basis tsunami descended on the Fukushima Daiichi Nuclear Power Plant by the earthquake. Eventually, the core cooling systems of the units 1, 2 and 3 could not operate stably, they all suffered severe accident, and hydrogen explosions were triggered in the reactor buildings of units 1, 3 and 4. In the light of these circumstances, Atomic Energy Society of Japan (AESJ) decided to establish a standard that consolidates the concept of maintaining and improving severe accident management. In the SAM standard, the combination of hardware and software measures based on the risk assessment enables a scientific and rational approach to apply to scenarios of various severe accidents including low-frequency, high-impact events, and assures safety with functionality and flexibility. The SAM standard is already established in March, 2014. After publication of the SAM standard, with regard to effectiveness assessment for accident management and treatment of the uncertainty of severe accident analysis code, for example, the detailed guideline will be prepared as appendices of the standard. (author)

  19. Methodological aspects to elaborate the management and procedure guides of severe accidents

    International Nuclear Information System (INIS)

    Gonzalez Gonzalez, F.; Jimenez Fernandez, A.

    1995-01-01

    The management guides in severe accidents are very important to know the procedures in these accidents. The present articles summarizes the methodological aspects to elaborate the management guides, in order to prevent the severe accidents

  20. Beyond Design Basis Severe Accident Management as an Element of DiD Concept Strengthening

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, M., E-mail: kuznetsov_mv@vosafety.ru [FSUE VO “Safety”, Moscow (Russian Federation)

    2014-10-15

    The 4{sup th} Level of DiD is ensured by management of beyond design basis accidents which is achieved by implementation of the Beyond Design Basis Accidents Management Guidance (BDBAMG) and, if necessary, by additional technical devices and organizational measures at NPP Unit. BDBAMG is located between Levels 3 and 5 in DiD and is related to them. It is connected with Level 3 by means of conditions generated at this Level and according to which BDBAM should be initiated (Level 4). It is associated with Level 5 by conditions which necessitate implementation of Emergency planning. Both types of conditions should be identified in BDBAMG. BDBAs including the phase of severe damage of fuel and protective barriers (severe accidents) in accordance with Russian regulatory framework are a subset of all BDBAs set. In this connection, such accident scenarios meet the representativeness criterion for further analysis and development of Guidance for their management. BDBAMG availability, as it provides robustness of DiD as a whole, is an obligatory condition for obtaining a NPP operational license. In the process of BDBAMG development and implementation a feedback with technical and organizational measures, comprising Level 1 and, to a less extent, Level 2, comes up. BDBAMG verification is an important final stage of its development. Addressing severe accidents, it is a challenging issue for a full scope simulator and may require its software modernization to make it responsive to severe accident phenomena. The existing BDBAMGs should be updated due to NPP Unit modernizations and in conjunction with the latest knowledge on severe accident phenomenology and lessons learnt from known events (e.g. NPP Fukushima). Thus, improvements incorporated in BDBAMG, enhance the strength of DiD. (author)

  1. 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

    framework of the MELCOR analysis will be helpful to establish a basis for the severe accident mitigation strategy. As shown in Table 5, all postulated major scenarios were estimated resulting in the reduction of water inventory level and the mass of hydrogen generation. In the case of loss of cooling accident, time to water level reduction was obviously dependent of the decay heat load. It was revealed that the loss of inventory was dominant factor in the case of loss of coolant accident. With respect to the mass of hydrogen generation, massive amount of hydrogen was generated from the molten corium concrete reaction in MELCOR calculation. Through the MELCOR modeling developed in this study, it was found that MELCOR program has capabilities to predict entire progress of water level, fuel degradation, hydrogen generation, and concrete reaction.

  2. Computational Aerodynamics of Shuttle Orbiter Damage Scenarios in Support of the Columbia Accident Investigation

    Science.gov (United States)

    Bibb, Karen L.; Prabhu, Ramadas K.

    2004-01-01

    In support of the Columbia Accident Investigation, inviscid computations of the aerodynamic characteristics for various Shuttle Orbiter damage scenarios were performed using the FELISA unstructured CFD solver. Computed delta aerodynamics were compared with the reconstructed delta aerodynamics in order to postulate a progression of damage through the flight trajectory. By performing computations at hypervelocity flight and CF4 tunnel conditions, a bridge was provided between wind tunnel testing in Langley's 20-Inch CF4 facility and the flight environment experienced by Columbia during re-entry. The rapid modeling capability of the unstructured methodology allowed the computational effort to keep pace with the wind tunnel and, at times, guide the wind tunnel efforts. These computations provided a detailed view of the flowfield characteristics and the contribution of orbiter components (such as the vertical tail and wing) to aerodynamic forces and moments that were unavailable from wind tunnel testing. The damage scenarios are grouped into three categories. Initially, single and multiple missing full RCC panels were analyzed to determine the effect of damage location and magnitude on the aerodynamics. Next is a series of cases with progressive damage, increasing in severity, in the region of RCC panel 9. The final group is a set of wing leading edge and windward surface deformations that model possible structural deformation of the wing skin due to internal heating of the wing structure. By matching the aerodynamics from selected damage scenarios to the reconstructed flight aerodynamics, a progression of damage that is consistent with the flight data, debris forensics, and wind tunnel data is postulated.

  3. Severe accident management at South Africa's Koeberg plant

    International Nuclear Information System (INIS)

    Prior, R.P.; Wolvaardt, F.P.; Holderbaum, D.F.; Lutz, R.J.; Taylor, J.J.; Hodgson, C.D.

    1997-01-01

    Between the middle of 1993 and the end of 1995, Westinghouse and Eskom implemented plant specific Severe Accident Management Guidelines (SAMGs) at the Koeberg Nuclear Power Plant in South Africa. Prior to this project, Koeberg, like many plants, had emergency operating procedures which contain guidance for plant personnel to perform preventive accident management measures in event of an accident. There was, however, no structured guidance on recovery from an event which progresses past core damage -mitigative accident management. The SAMGs meet this need. In this paper, the Westinghouse approach to severe accident management is outlined, and the Koeberg implementation project described. A few key issues which arose during implementation are discussed, including plant instrumentation, flooding of the reactor pit, organisation and training of the Technical Support Centre staff, and impact of SAMG on risk. The means by which both generic and plant-specific SAMG have been validated is also summarised. In the next few years, many LWR owners will be implementing SAMG. In the U.S. all plants are in the process of developing SAMG. The Koeberg project is believed to be the first plant specific implementation of the WOG SAMG worldwide, and this paper has hopefully provided insights into some of the implementation issues for those about to undertake similar projects. (author)

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

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-12-01

    The work Programme of the ISARRP Project was divided into several work packages. The work was conducted in the form of presentations and discussions, held during several meetings whose character was that of workshops. Short reports were prepared by the partners assigned to each task. Work Package 1: Critical review of the SA phenomenological research. The objective of this work package was to consider the progress made world-wide in research on the resolution of the outstanding phenomenological issues posed by severe accidents. Work Package 2: Relevance of severe accident research to SAMG requirements and implementation. The objective of this work package was to relate the progress made in the resolution of the SA issues to the practical matter of what results are required or have been used for the management of severe accidents. Clearly, the SAMG is the most important avenue employed by the regulatory organizations to assure themselves of the safe (from public perspective) performance of a nuclear plant in a postulated severe accident event. Work Package 3: Relevance of severe accident research to PSA and the risk informed regulatory approach. The objectives of this work package is to relate the results obtained by the severe accident research to the requirements of a PSA and of the new trend of employing the risk informed approach in promulgating regulations. Clearly a PSA identifies vulnerabilities in the knowledge base, however, their importance is decidedly plant specific. Nevertheless the uncertainties in the phenomenology or in resolution of issues lead to uncertainties in the PSA conclusions and in the adoption of the risk informed approach. Work Package 4: Questionnaire and the evaluation of responses to the questions. The purpose of this work package is to solicit the views of the regulatory organizations towards the results of the SA research and the benefits they have derived from it in terms of regulatory actions, or in the confidence they have gained

  5. 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

    The work Programme of the ISARRP Project was divided into several work packages. The work was conducted in the form of presentations and discussions, held during several meetings whose character was that of workshops. Short reports were prepared by the partners assigned to each task. Work Package 1: Critical review of the SA phenomenological research. The objective of this work package was to consider the progress made world-wide in research on the resolution of the outstanding phenomenological issues posed by severe accidents. Work Package 2: Relevance of severe accident research to SAMG requirements and implementation. The objective of this work package was to relate the progress made in the resolution of the SA issues to the practical matter of what results are required or have been used for the management of severe accidents. Clearly, the SAMG is the most important avenue employed by the regulatory organizations to assure themselves of the safe (from public perspective) performance of a nuclear plant in a postulated severe accident event. Work Package 3: Relevance of severe accident research to PSA and the risk informed regulatory approach. The objectives of this work package is to relate the results obtained by the severe accident research to the requirements of a PSA and of the new trend of employing the risk informed approach in promulgating regulations. Clearly a PSA identifies vulnerabilities in the knowledge base, however, their importance is decidedly plant specific. Nevertheless the uncertainties in the phenomenology or in resolution of issues lead to uncertainties in the PSA conclusions and in the adoption of the risk informed approach. Work Package 4: Questionnaire and the evaluation of responses to the questions. The purpose of this work package is to solicit the views of the regulatory organizations towards the results of the SA research and the benefits they have derived from it in terms of regulatory actions, or in the confidence they have gained

  6. Neural Correlates of Posttraumatic Growth after Severe Motor Vehicle Accidents

    Science.gov (United States)

    Rabe, Sirko; Zollner, Tanja; Maercker, Andreas; Karl, Anke

    2006-01-01

    Frontal brain asymmetry has been associated with emotion- and motivation-related constructs. The authors examined the relationship between frontal brain asymmetry and subjective perception of posttraumatic growth (PTG) after severe motor vehicle accidents (MVAs). Eighty-two survivors of MVAs completed self-report measures of PTG, trait and state…

  7. 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.

  8. Severe Accident R and D for Enhanced CANDU-6 Reactors

    International Nuclear Information System (INIS)

    Nitheanandan, Thambiayah

    2012-01-01

    CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

  9. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

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

    2007-06-01

    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

  10. Strategies to cope with severe accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Kovacs, Zoltan; Rydzi, Stanislav

    2015-01-01

    The paper focusses, in particular, on SAMG – Severe Accident Management Guidelines, and on SBEOP - Symptom Based Emergency Operating Procedures. It is shown how the concepts are applicable, how they are applied in practice, and in which aspects they need improvements. (orig.)

  11. Design study on dose evaluation method for employees at severe accident

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio; Kudo, Seiichi; Nishimura, Kazuya

    2002-01-01

    If a severe accident occurs in a pressurized water reactor plant, it is required to estimate dose values of operators engaged in emergency such as accident management, repair of failed parts. However, it might be difficult to measure radiation dose rate during the progress of an accident, because radiation monitors are not always installed in areas where the emergency activities are required. In this study, we analyzed the transport of radioactive materials in case of a severe accident, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system from this design study, and then evaluated its availability. As a result, we obtained the following: (1) A new dose evaluation method was established to predict the radiation dose rate at any point in the plant during a severe accident scenario. (2) This evaluation of total dose including access route and time for emergency activities is useful for estimating radiation dose limit for these employee actions. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  12. Mitigation of severe accidents in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Soederman, E.

    1987-01-01

    Sweden is the first country to build filtered venting systems, the first one became operable at Barsebaeck nuclear power plant in 1985. In new concepts, now being installed in Sweden, an enhanced containment spray system is the basic element and the filtered venting is only the secondary mitigating system. The filter is a new design, a submerged multi venturi scrubber. The Swedish strategy has been built on three basics: improved knowledge through research; containment integrity through mitigating systems; and accident management to prevent severe accidents. 2 figs

  13. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

    Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  15. Considerations on monitoring needs of advanced, passive safety light water reactors for severe accident management

    International Nuclear Information System (INIS)

    Bava, G.; Zambardi, F.

    1992-01-01

    This paper deals with problems concerning information and related instrumentation needs for Accident Management (AM), with special emphasis on Severe Accidents (SA) in the new advanced, passive safety Light Water Reactors (PLWR), presently in a development stage. The passive safety conception adopted in the plants concerned goes parallel with a deeper consideration of SA, that reflects the need of increasing the plant resistance against conditions going beyond traditional ''design basis accidents''. Further, the role of Accident Management (AM) is still emphasized as last step of the defence in depth concept, in spite of the design efforts aimed to reduce human factor importance; as a consequence, the availability of pertinent information on actual plant conditions remains a necessary premise for performing preplanned actions. This information is essential to assess the evolution of the accident scenarios, to monitor the performances of the safety systems, to evaluate the ultimate challenge to the plant safety, and to implement the emergency operating procedures and the emergency plans. Based on these general purposes, the impact of the new conception on the monitoring structure is discussed, furthermore reference is made to the accident monitoring criteria applied in current plants to evaluate the requirements for possible solutions. (orig.)

  16. Temporary jobs and the severity of workplace accidents.

    Science.gov (United States)

    Picchio, Matteo; van Ours, Jan C

    2017-06-01

    From the point of view of workplace safety, it is important to know whether having a temporary job has an effect on the severity of workplace accidents. We present an empirical analysis on the severity of workplace accidents by type of contract. We used microdata collected by the Italian national institute managing the mandatory insurance against work related accidents. We estimated linear models for a measure of the severity of the workplace accident. We controlled for time-invariant fixed effects at worker and firm levels to disentangle the impact of the type of contract from the spurious one induced by unobservables at worker and firm levels. Workers with a temporary contract, if subject to a workplace accident, were more likely to be confronted with severe injuries than permanent workers. When correcting the statistical analysis for injury under-reporting of temporary workers, we found that most of, but not all, the effect is driven by the under-reporting bias. The effect of temporary contracts on the injury severity survived the inclusion of worker and firm fixed effects and the correction for temporary workers' injury under-reporting. This, however, does not exclude the possibility that, within firms, the nature of the work may vary between different categories of workers. For example, temporary workers might be more likely to be assigned dangerous tasks because they might have less bargaining power. The findings will help in designing public policy effective in increasing temporary workers' safety at work and limiting their injury under-reporting. Copyright © 2017. Published by Elsevier Ltd.

  17. 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).

  18. External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

    International Nuclear Information System (INIS)

    Mihalina, M.; Spalj, S.; Glaser, B.

    2016-01-01

    The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

  19. Remediation strategies after nuclear or radiological accidents: part 2 - accident scenarios for assessing effectiveness of cleanup procedures

    International Nuclear Information System (INIS)

    Rochedo, Elaine R.R.

    2009-01-01

    The selection of protective measures and remediation strategies after an accident needs to be based on previously established criteria, to minimize unnecessary stress and the exposures involved in cleanup operations that are not effective in reducing doses to the public. In a first stage, a database describing the countermeasures has been developed including their efficiency on removing contamination from surfaces. However, to assess the effectiveness of cleanup procedures in reducing doses to members of the public, it was necessary to derive specific scenarios in order to simulate the long term behavior of the material in the environment, since the contribution of different surfaces to doses changes with time after contamination. A basic release and exposure scenario was developed to assess the dose reduction due to the mostly used procedures. Exposure scenarios were selected to fit the surroundings of the Brazilian nuclear power plants in Angra dos Reis. Simulations were performed using SIEM, the integrated system for dose assessment after contamination events, developed at IRD. The contamination of urban environments was assessed for Cs-137, as this was found to be the most relevant long term radionuclide to contribute to doses to member of the public. The effects on reducing external exposures were assessed for periods up to 50 years after the contamination. For agricultural areas, the focus was on ingestion doses from contamination with I-131 for periods up to 1 year after contamination. Results will be complemented on the database in order to support multi-criteria decision making processes after accidents. (author)

  20. Remediation strategies after nuclear or radiological accidents: part 2 - accident scenarios for assessing effectiveness of cleanup procedures

    Energy Technology Data Exchange (ETDEWEB)

    Rochedo, Elaine R.R. [Comissao Nacional de Energia Nuclear (CNEN-RJ), Rio de Janeiro, RJ (Brazil). Coordenacao de Instalacoes Nucleares], e-mail: erochedo@cnen.gov.br; Silva, Diogo N.G.; Wasserman, Maria A.V.; Conti, Luiz F.C. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)], e-mail: dneves@ird.gov.br, e-mail: angelica@ird.gov.br, e-mail: lfcconti@ird.gov.br

    2009-07-01

    The selection of protective measures and remediation strategies after an accident needs to be based on previously established criteria, to minimize unnecessary stress and the exposures involved in cleanup operations that are not effective in reducing doses to the public. In a first stage, a database describing the countermeasures has been developed including their efficiency on removing contamination from surfaces. However, to assess the effectiveness of cleanup procedures in reducing doses to members of the public, it was necessary to derive specific scenarios in order to simulate the long term behavior of the material in the environment, since the contribution of different surfaces to doses changes with time after contamination. A basic release and exposure scenario was developed to assess the dose reduction due to the mostly used procedures. Exposure scenarios were selected to fit the surroundings of the Brazilian nuclear power plants in Angra dos Reis. Simulations were performed using SIEM, the integrated system for dose assessment after contamination events, developed at IRD. The contamination of urban environments was assessed for Cs-137, as this was found to be the most relevant long term radionuclide to contribute to doses to member of the public. The effects on reducing external exposures were assessed for periods up to 50 years after the contamination. For agricultural areas, the focus was on ingestion doses from contamination with I-131 for periods up to 1 year after contamination. Results will be complemented on the database in order to support multi-criteria decision making processes after accidents. (author)

  1. 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.

  2. Evaluation of severe accident risks, Grand Gulf, Unit 1: Appendices

    International Nuclear Information System (INIS)

    Brown, T.D.; Breeding, R.J.; Jow, H.N.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.; Amos, C.N.

    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 report in 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 Grand Gulf Nuclear Station, Unit 1. This power plant, located in Port Gibson, Mississippi, is operated by the System Energy Resources, Inc. (SERI). 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 internal to the power plant was assessed. This document provides Appendices A through E for this report. Topics included are, respectively: supporting information for the accident progression analysis; supporting information for the source term analysis; supporting information for the consequence analysis; risk results; and sampling information

  3. ANS severe accident program overview ampersand planning document

    International Nuclear Information System (INIS)

    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 -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

  4. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  5. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    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 is 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 is 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 is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  6. Comparative risk assessment of severe accidents in the energy sector

    International Nuclear Information System (INIS)

    Burgherr, Peter; Hirschberg, Stefan

    2014-01-01

    Comparative assessment of accident risks in the energy sector is a key aspect in a comprehensive evaluation of sustainability and energy security concerns. Safety performance of energy systems can have important implications on the environmental, economic and social dimensions of sustainability as well as availability, acceptability and accessibility aspects of energy security. Therefore, this study provides a broad comparison of energy technologies based on the objective expression of accident risks for complete energy chains. For fossil chains and hydropower the extensive historical experience available in PSI's Energy-related Severe Accident Database (ENSAD) is used, whereas for nuclear a simplified probabilistic safety assessment (PSA) is applied, and evaluations of new renewables are based on a combination of available data, modeling, and expert judgment. Generally, OECD and EU 27 countries perform better than non-OECD. Fatality rates are lowest for Western hydropower and nuclear as well as for new renewables. In contrast, maximum consequences can be by far highest for nuclear and hydro, intermediate for fossil, and very small for new renewables, which are less prone to severe accidents. Centralized, low-carbon technology options could generally contribute to achieve large reductions in CO 2 -emissions; however, the principal challenge for both fossil with Carbon Capture and Storage and nuclear is public acceptance. Although, external costs of severe accidents are significantly smaller than those caused by air pollution, accidents can have disastrous and long-term impacts. Overall, no technology performs best or worst in all respects, thus tradeoffs and priorities are needed to balance the conflicting objectives such as energy security, sustainability and risk aversion to support rationale decision making. - Highlights: • Accident risks are compared across a broad range of energy technologies. • Analysis of historical experience was based on the

  7. Some outstanding issues in severe accidents containment performance

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2004-01-01

    This paper describes the current status of the outstanding issues in severe accident performance of Light Water Reactor containments that have been raised in the last several years. The results of the research that has been performed on the topics concerning these issues will be described. Some of these issues have been resolved, some are close to resolution, while others need further evaluation and research results. (author)

  8. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  9. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

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

    International Nuclear Information System (INIS)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu

    2017-01-01

    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

  11. 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.

  12. Proceedings of the workshop on operator training for severe accident management and instrumentation capabilities during severe accidents

    International Nuclear Information System (INIS)

    2001-01-01

    This Workshop was organised in collaboration with Electricite de France (Service Etudes et Projets Thermiques et Nucleaires). There were 34 participants, representing thirteen OECD Member countries, the Russian Federation and the OECD/NEA. Almost half the participants represented utilities. The second largest group was regulatory authorities and their technical support organisations. Basically, the Workshop was a follow-up to the 1997 Second Specialist Meeting on Operator Aids for Severe Accident Management (SAMOA-2) [Reports NEA/CSNI/R(97)10 and 27] and to the 1992 Specialist Meeting on Instrumentation to Manage Severe Accidents [Reports NEA/CSNI/R(92)11 and (93)3]. It was aimed at sharing and comparing progress made and experience gained from these two meetings, emphasizing practical lessons learnt during training or incidents as well as feedback from instrumentation capability assessment. The objectives of the Workshop were therefore: - to exchange information on recent and current activities in the area of operator training for SAM, and lessons learnt during the management of real incidents ('operator' is defined hear as all personnel involved in SAM); - to compare capabilities and use of instrumentation available during severe accidents; - to monitor progress made; - to identify and discuss differences between approaches relevant to reactor safety; - and to make recommendations to the Working Group on the Analysis and Management of Accidents and the CSNI (GAMA). The meeting confirmed that only limited information is needed for making required decisions for SAM. In most cases existing instrumentation should be able to provide usable information. Additional instrumentation requirements may arise from particular accident management measures implemented in some plants. In any case, depending on the time frame where the instrumentation should be relied upon, it should be assessed whether it is likely to survive the harsh environmental conditions it will be exposed

  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. A methodology for analysing human errors of commission in accident scenarios for risk assessment

    International Nuclear Information System (INIS)

    Kim, J. H.; Jung, W. D.; Park, J. K

    2003-01-01

    As the concern on the impact of the operator's inappropriate interventions, so-called Errors Of Commissions(EOCs), on the plant safety has been raised, the interest in the identification and analysis of EOC events from the risk assessment perspective becomes increasing accordingly. To this purpose, we propose a new methodology for identifying and analysing human errors of commission that might be caused from the failures in situation assessment and decision making during accident progressions given an initiating event. The proposed methodology was applied to the accident scenarios of YGN 3 and 4 NPPs, which resulted in about 10 EOC situations that need careful attention

  15. Assessment and comparison of two early warning indicator methods in the perspective of prevention of atypical accident scenarios

    International Nuclear Information System (INIS)

    Paltrinieri, Nicola; Øien, Knut; Cozzani, Valerio

    2012-01-01

    Some severe major accidents occurred in Europe in recent years (e.g. the Vapour Cloud Explosion at Buncefield in 2005), which were not foreseen by their site “Seveso-II” safety reports. Detailed analyses of such “atypical” scenarios demonstrated that they are the result of a number of failures at different technical and organizational levels. Thus, their prevention is a major challenge and must be coordinated through different kinds of approaches, among which improved early detection plays an important role. Proactive methodologies for the development of early warning indicators can unveil early deviations in the causal chain. Two examples are the Resilience-based Early Warning Indicator (REWI) method and the so-called “Dual Assurance” method. The aim of this study was to analyse the possible integration of early warning indicators in the hazard identification process. A Buncefield-like site was analysed to obtain indicators that were compared with the actual causes that led to the accident at Buncefield (and to similar accident scenarios). The results show that indicators from both methods could have prevented the accidents from happening. However, one main difference is related to the issue of hazard identification, which is fundamental for the prevention of atypical accident scenarios. The REWI method is not dependent on the outcome of the hazard identification process. Instead it provides complementarities to the first prevention approach (improved identification of atypical scenarios), demonstrating that a mutual activity would be an effective strategy in which human, organizational, cultural and technical factors are treated in an integrated manner. - Highlights: ► Early warning indicators were created through 2 methods for the Buncefield oil depot. ► A general capacity to cover causes of atypical events was demonstrated. ► The Dual Assurance method showed to mainly cover operability failures. ► The REWI method showed to promote acts

  16. Summary and conclusions: Specialist Meeting on Severe Accident Management Implementation

    International Nuclear Information System (INIS)

    1995-01-01

    During the first session of this meeting, regulators, research groups, designers/owners' groups and some utilities discussed the critical decisions in SAM (Severe Accident Management), how these decisions were addressed and implemented in generic SAM guidelines, what equipment and instrumentation was used, what are the differences in national approaches, etc. During the second session, papers were presented by utility specialists that described approaches chosen for specific implementation of the generic guidelines, the difficulties encountered in the implementation process and the perceived likelihood of success of their SAM programme in dealing with severe accidents. The third and final sessions was dedicated to discussing what are the remaining uncertainties and open questions in SAM. Experts from several OECD countries presented significant perspectives on remaining open issues

  17. Proceedings of the specialist meeting on severe accident management implementation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Niantic Specialist meeting was structured around three main themes, one for each session. During the first session, papers from regulators, research groups, designers/owners groups and some utilities discussed the critical decisions in Severe Accident Management (SAM), how these decisions were addressed and implemented in generic SAM guidelines, what equipment and instrumentation was used, what are the differences in national approaches, etc. During the second session, papers were presented by utility specialists that described approaches chosen to specific implementation of the generic guidelines, the difficulties encountered in the implementation process and the perceived likelihood of success of their SAM program in dealing with severe accidents. The third session was dedicated to discussing what are the remaining uncertainties and open questions in SAM. Experts from several OECD countries presented significant perspectives on remaining open issues

  18. Proceedings of the specialist meeting on severe accident management implementation

    International Nuclear Information System (INIS)

    1995-01-01

    The Niantic Specialist meeting was structured around three main themes, one for each session. During the first session, papers from regulators, research groups, designers/owners groups and some utilities discussed the critical decisions in Severe Accident Management (SAM), how these decisions were addressed and implemented in generic SAM guidelines, what equipment and instrumentation was used, what are the differences in national approaches, etc. During the second session, papers were presented by utility specialists that described approaches chosen to specific implementation of the generic guidelines, the difficulties encountered in the implementation process and the perceived likelihood of success of their SAM program in dealing with severe accidents. The third session was dedicated to discussing what are the remaining uncertainties and open questions in SAM. Experts from several OECD countries presented significant perspectives on remaining open issues

  19. Pending issues for severe accident management in Wolsong plants

    International Nuclear Information System (INIS)

    Song, Y.M.; Kim, D.H.; Park, S.Y.

    2015-01-01

    While the fraction of electric power supplied from a PHWR is more than 10% in Korea, the establishment of PHWR safety enhancement based on the SAM (Severe Accident Management) technology is still weak. The final approval on the extended operation and a stress test of Wolsong-1 were made under the condition that SAM is to be enhanced. Under this situation, the current research at KAERI of Korea has a vision to strengthen the unique value of a PHWR by resolving the pending SAM issues devaluating the PHWRs’ original value. Research activities in this area will be presented. This presentation will include: The operating strategy of CFVS (Containment Filtered Vent System) for Wolsong in which vent size and closure pressure are treated because some peak spikes (at failure times of calandria and calandria vault) are difficult to be controlled; Reactor Building failure pressure at which failure probability is treated for different modes such as global and leak failures; the adequacy of DCRV (Degasser Condenser tank Relief Valve) steam relief capacity with severe SGTR source term, and Hydrogen generation and control issue which is specific to CANDU. Furthermore, current SAM guidance has a lack of information on accident diagnostic and prognostic analyses, which is difficult for the TSC (Technical Service Center) emergency staff members to deal with under real accident conditions. Thus, prototypic technologies (such as an accident inferring engine and simulator) together with SAM updates are being developed as key elements to SAM supporting tools called SAMEX-CANDU

  20. A PC based multi-CPU severe accident simulation trainer

    International Nuclear Information System (INIS)

    Jankowski, M.W.; Bienarz, P.P.; Sartmadjiev, A.D.

    2004-01-01

    MELSIM Severe Accident Simulation Trainer is a personal computer based system being developed by the International Atomic Energy Agency and Risk Management Associates, Inc. for the purpose of training the operators of nuclear power stations. It also serves for evaluating accident management strategies as well as assessing complex interfaces between emergency operating procedures and accident management guidelines. The system is being developed for the Soviet designed WWER-440/Model 213 reactor and it is plant specific. The Bohunice V2 power station in the Slovak Republic has been selected for trial operation of the system. The trainer utilizes several CPUs working simultaneously on different areas of simulation. Detailed plant operation displays are provided on colour monitor mimic screens which show changing plant conditions in approximate real-time. Up to 28 000 curves can be plotted on a separate monitor as the MELSIM program proceeds. These plots proceed concurrently with the program, and time specific segments can be recalled for review. A benchmarking (limited in scope) against well validated thermal-hydraulic codes and selected plant accident data (WWER-440/213 Rovno NPP, Ukraine) has been initiated. Preliminary results are presented and discussed. (author)

  1. Insights from Severe Accident Analyses for Verification of VVER SAMG

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

  2. Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

    2012-01-01

    Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

  3. Shipping container response to severe highway and railway accident conditions: Appendices

    International Nuclear Information System (INIS)

    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. Comparison of risk-dominant scenario assumptions for several TRU waste facilities in the DOE complex

    International Nuclear Information System (INIS)

    Foppe, T.L.; Marx, D.R.

    1999-01-01

    In order to gain a risk management perspective, the DOE Rocky Flats Field Office (RFFO) initiated a survey of other DOE sites regarding risks from potential accidents associated with transuranic (TRU) storage and/or processing facilities. Recently-approved authorization basis documents at the Rocky Flats Environmental Technology Site (RFETS) have been based on the DOE Standard 3011 risk assessment methodology with three qualitative estimates of frequency of occurrence and quantitative estimates of radiological consequences to the collocated worker and the public binned into three severity levels. Risk Class 1 and 2 events after application of controls to prevent or mitigate the accident are designated as risk-dominant scenarios. Accident Evaluation Guidelines for selection of Technical Safety Requirements (TSRs) are based on the frequency and consequence bin assignments to identify controls that can be credited to reduce risk to Risk Class 3 or 4, or that are credited for Risk Class 1 and 2 scenarios that cannot be further reduced. This methodology resulted in several risk-dominant scenarios for either the collocated worker or the public that warranted consideration on whether additional controls should be implemented. RFFO requested the survey because of these high estimates of risks that are primarily due to design characteristics of RFETS TRU waste facilities (i.e., Butler-type buildings without a ventilation and filtration system, and a relatively short distance to the Site boundary). Accident analysis methodologies and key assumptions are being compared for the DOE sites responding to the survey. This includes type of accidents that are risk dominant (e.g., drum explosion, material handling breach, fires, natural phenomena, external events, etc.), source term evaluation (e.g., radionuclide material-at-risk, chemical and physical form, damage ratio, airborne release fraction, respirable fraction, leakpath factors), dispersion analysis (e.g., meteorological

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

    International Nuclear Information System (INIS)

    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

  6. Numerical module for debris behavior under severe accident conditions

    International Nuclear Information System (INIS)

    Kisselev, A.E.; Kobelev, G.V.; Strizhov, V.F.; Vasiliev, A.D.

    2005-01-01

    The late phase of a hypothetical severe accident in a nuclear reactor is characterized by the appearance of porous debris and liquid pools in core region and lower head of the reactor vessel. Thermal hydraulics and heat transfer in these regions are very important for adequate analysis of severe accident dynamics. The purpose of this work is to develop a universal module which is able to model above-mentioned phenomena on the basis of modern physical concepts. The original approach for debris evolution is developed from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The calculation results of several tests on modeling of porous debris behavior, including the MP-1 experiment, are presented in comparison with experimental data. The results are obtained using this module implemented into the Russian best estimate code, RATEG/SVECHA/HEFEST, which was developed for modeling severe accident thermal hydraulics and late phase phenomena in VVER nuclear power plants. (author)

  7. Overview of severe accident research at the USNRC

    International Nuclear Information System (INIS)

    Basu, S.; Ader, C.E.

    1999-01-01

    This paper summarizes the U.S. Nuclear Regulatory Commission's (USNRC) severe accident research activities, in particular, progress made in the past year toward the resolution and/or improved understanding of a number of severe accident issues. The direct containment heating (DCH) is nearing resolution for Combustion Engineering and Babcock and Wilcox type pressurized water reactors (PWRs) are well as for ice condensers. Additionally, two lower pressure DCH tests were conducted recently at the Sandia National Laboratories (SNL) under the NRC/IPSN/FzK sponsorship to provide data regarding intentional depressurization as an accident management strategy to mitigate DCH loads. In the area of lower head integrity, the experimental program to investigate boiling heat transfer on downward facing curved surfaces with insulation was completed. Finally, the SNL program investigating the creep rupture behavior of the lower head under the combined thermo-mechanical loading was completed recently. Additional lower head experiments at SNL are being planned as an OECD project. During the past year, the USNRC participated in two programs aimed at extending the data base on hydrogen combustion into more prototypic situations. Testing was performed at the Brookhaven National Laboratory (BNL) to investigate detonation transmission at elevated temperatures. In a cooperative program under the sponsorship of NRC/IPSN/FzK, Russian Research Center (RRC) investigated hydrogen combustion issues at large scale at the RUT facility. The experimental program at the SNL to examine the performance of Passive Autocatalytic Recombiners (PARs) was completed also this year. In the fuel-coolant interaction (FCI) area, the experimental work at the Argonne National Laboratory (ANL) to investigate chemical augmentation of FCI energetics was completed as was the experimental work at the University of Wisconsin (UW) involving one-dimensional propagation experiments (similar to KROTOS). The USNRC is

  8. Safety against releases in severe accidents. Final report

    International Nuclear Information System (INIS)

    Lindholm, I.; Berg, Oe.; Nonboel, E.

    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)

  9. 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.

  10. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  11. ESTER: a new approach in modelling severe accidents

    International Nuclear Information System (INIS)

    Shepherd, I.; Jones, A.; Schmidt, F.

    1993-01-01

    ESTER is a set of codes for calculating phenomena during severe accidents in thermal reactors. It makes use of software tools that allow the data to be defined as a tree-structured data base and this data to be stored and retrieved by the code modules. The tools include generalized input and output routines that are independent of the particular code being used. Severe accident research codes are in a continual state of development and the structure of ESTER is such that modifications can be introduced easily and safely. The ESTER framework also facilitates the coupling together of codes. A preliminary version of ESTER containing a complete set of tools and a limited number of applications has already been released. 9 refs., 5 figs

  12. Vessel-related problems in severe accidents, International Research Projects

    International Nuclear Information System (INIS)

    Figueras, J. M.

    2000-01-01

    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)

  13. Numerical simulation of gasket behaviour during severe accidents (ATHERMIP project)

    International Nuclear Information System (INIS)

    Castro Lopez, Fernando; Orden Martinez, Alfredo

    1998-01-01

    This paper summarises the work carried out to numerically simulate the thermo-mechanical behaviour of sealing gasket in large containment penetrations during a severe accident. The gasket material is an elastomeric material and the thermo-mechanical characterization was based on experimentation. The difficulty of numerical simulation lies in the high non-linearity of the analysis, due on one hand, to the high strain levels reached, and on the other, to stiffness changes introduced by contact/takeoff indicators. Also, the stiffness parameters of the gasket material are not constant, but are subject to changes, both regarding the strain level and the environmental conditions (temperature, radiation). The results obtained allow presenting a calculation model capable of simulating and explaining the behaviour of the sealing gasket during a severe accident. Also, the failure hypothesis numerically obtained was environmentally validated. (author)

  14. Severe accidents risk assessment as a basis for emergency preparedness

    International Nuclear Information System (INIS)

    Sinka, D.; Mikulicic, V.

    2000-01-01

    The paper demonstrates, by example of the Republic of Croatia, the possibilities of implementing risk assessment as basis for nuclear accident emergency preparedness development. Individual risks of severe accidents for citizens of the biggest Croatian population centers, as well as collective risk for entire population have been assessed using the PRONEL method. The assessment covered 90 power reactors located at a distance up to 1.000 km. The conducted assessment shows the risks for various regions of the Republic of Croatia, and comparison between them. If risk would be taken as basic criterion in nuclear emergency planning, the results of assessment would directly indicate the necessary preparation level for each region. Furthermore, the assessment of risks from individual power plants and power plant types indicates to which facilities the greatest attention should be paid in nuclear accidents preparedness development. Risks from groups of power plants formed in accordance with their respective distance from exposure location shows what kind of tools for determining consequences and protective actions during a nuclear accident should be made available. (author)

  15. MAAP - modular program for analyses of severe accidents

    International Nuclear Information System (INIS)

    Henry, R.E.; Lutz, R.J.

    1990-01-01

    The MAAP computer code was developed by Westinghouse as a fast, user-friendly, integrated analytical tool for evaluations of the sequences and consequences of severe accidents. The code allows a fully integrated treatment of thermohydraulic behavior and of the fission products in the primary system, the containment, and the ancillary buildings. This ensures interactive inclusion of all thermohydraulic events and of fission product behavior. All important phenomena which may occur in a major accident are contained in the modular code. In addition, many of the important parameters affecting the multitude of different phenomena can be defined by the user. In this way, it is possible to study the accuracy of the predicted course and of the consequences of a series of major accident phenomena. The MAAP code was subjected to extensive benchmarking with respect to the results of the experimental and theoretical programs, the findings obtained in other safety analyses using computers and data from accidents and transients in plants actually in operation. With the expected connection of the validation and test programs, the computer code attains a quality standard meeting the most stringent requirements in safety analyses. The code will be enlarged further in order to expand the number of benchmarks and the resolution of individual comparisons, and to ensure that future MAAP models will be in better agreement with the experiments and experiences of industry. (orig.) [de

  16. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Cahalan, J.E.

    2009-01-01

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

  17. Molten Corium-Concrete Interaction Behavior Analyses for Severe Accident Management in CANDU Reactor

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, D. H.; Song, Y. M.

    2014-01-01

    After the last few severe accidents, the importance of accident management in nuclear power plants has increased. Many countries, including the United States (US) and Canada, have focused on understanding severe accidents in order to identify ways to further improve the safety of nuclear plants. It has been recognized that severe accident analyses of nuclear power plants will be beneficial in understanding plant-specific vulnerabilities during severe accidents. The objectives of this paper are to describe the molten corium behavior to identify a plant response with various concrete specific components. Accident analyses techniques using ISSAC can be useful tools for MCCI behavior in severe accident mitigation

  18. Severe accident prevention and mitigation: A utility perspective - EDF approach

    International Nuclear Information System (INIS)

    Vidard, M.

    1998-01-01

    Current plans have excellent safety records and are cost competitive. For future plants, excellence in safety will remain a prerequisite, as well as increased cost competitiveness. When contemplating solutions to Severe Accident challenges, cost effectiveness is essential in the decision making process. This cost effectiveness must be understood not only in terms of capital cost, but also of Operation and Maintenance costs as well as absence of additional risks to plant operators. Examples are given to illustrate the recommended approach

  19. Containment hydrogen and atmosphere activity control to mitigate severe accidents in VVERs and Western PWRs. Design and status of implementation

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2002-01-01

    For accident management nuclear power plants in Europe have been or will be back-fitted with supplementary systems for monitoring the containment hydrogen concentration, for the early removal and reduction of hydrogen and filtered venting systems to retain radioactive aerosols and iodine. The hydrogen monitoring system (HMS) provides the information of local H 2 concentration in the containment during DBA and severe accident situations. The new HMS contains of overall H 2 -sensors and is installed inside the confinement. It provides continuos information about the local and temporal distribution of hydrogen, reported directly to the Emergency Response Team in case of severe accident. The hydrogen Reduction System (HRS) consists of several Passive Autocatalytic Recombiners (PAR) located in several compartments in the containment. The number of PARs to be installed depends on the type of NPP, structure of containment and the investigated accident scenario e.g. DBA conditions - approx. 6 to 20 PARs; severe accident conditions - 20-60 PARs). In case of severe accident it does not need any operator actions. The Filtered Venting System (FVS) is is especially important for WWER-440/230 maintaining sub atmospheric pressure in the confinement. For severe accident the on-site Emergency Response Team has to take the necessary strategic decisions for containment depressurization via the FVS

  20. 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)

  1. Regulatory analyses for severe accident issues: an example

    International Nuclear Information System (INIS)

    Burke, R.P.; Strip, D.R.; Aldrich, D.C.

    1984-09-01

    This report presents the results of an effort to develop a regulatory analysis methodology and presentation format to provide information for regulatory decision-making related to severe accident issues. Insights and conclusions gained from an example analysis are presented. The example analysis draws upon information generated in several previous and current NRC research programs (the Severe Accident Risk Reduction Program (SARRP), Accident Sequence Evaluation Program (ASEP), Value-Impact Handbook, Economic Risk Analyses, and studies of Vented Containment Systems and Alternative Decay Heat Removal Systems) to perform preliminary value-impact analyses on the installation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The results presented in this report are first-cut estimates, and are presented only for illustrative purposes in the context of this document. This study should serve to focus discussion on issues relating to the type of information, the appropriate level of detail, and the presentation format which would make a regulatory analysis most useful in the decisionmaking process

  2. A Study on the Requisite Information for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunhee; Ahn, Kwang-Il; Kim, Jae-Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Related this research on arranging the requisite information for severe accident management, the documents of various forms in each country as well as the domestic literature are secured and analyzed. The analyzed information is arranged up to a detailed level. For the secured documents, the issued organizations and the issued purpose are diverse. Thus, the contents of the secured documents are also diverse according to the reactor type, and the purpose and standards of the classification are also diverse. Moreover, terminologies with same meaning are not unified. These various documents are analyzed to arrange the requisite information for severe accident management. Based on the documents of a related severe accident, the major information was analyzed. The information is different according to the reactor type, classification standard, and classification standard of the safety function. Thus the information is classified variously. In this study, based on the analysis results of the documents described these information, the major information and parameters are examined as safety function. And the results of parameters and information including the safety function and the detail information are induced.

  3. Influence of radiation heat transfer during a severe accident

    International Nuclear Information System (INIS)

    Cazares R, R. I.; Epinosa P, G.; Varela H, J. R.; Vazquez R, A.; Polo L, M. A.

    2016-09-01

    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)

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

    International Nuclear Information System (INIS)

    Lee Min; Ko, Y.-C.

    2008-01-01

    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

  5. Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

    International Nuclear Information System (INIS)

    Petkov, Gueorgui

    2010-01-01

    Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

  6. Discussion on several issues of the accidents management of nuclear power plants in operation

    International Nuclear Information System (INIS)

    Cao Xuewu; Wang Zhe; Zhang Yingzhen

    2009-01-01

    This article discusses several issues of the accident management of nuclear power plants in operation, for example: the necessity, implementation principle of accident management and accident management program etc. For conducting accident management for beyond design basis accidents, this article thinks that the accident management program should be developed and implemented to ensure that the plant and its personnel with responsibilities for accident management are adequately prepared to take effective on-site actions to prevent or mitigate the consequences of severe accident. (authors)

  7. 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.

  8. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Fischer, M.; Bittermann, D.

    2005-01-01

    The EPR, an evolutionary pressurized water reactor (PWR), is a 4300-4500 MWth that incorporates proven technology within an optimized configuration to enhance safety. EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product. Commercial EPR units are currently being built in Finland at the Olkiluoto site, and planned for France at the Flamanville site. In recent months, Framatome ANP announced their intention to market the EPR units to China in response to a request for vendor bids as well as their intent to pursue design certification in the United States under 10CFR52. The EPR safety philosophy is based on a deterministic consideration of defense-in-depth complemented by probabilistic analyses. Not only is the EPR designed to prevent and mitigate design basis accidents (DBAs), it employs an extra level of safety associated with severe accident response. Therefore, as a design objective, features are included to ensure that radiological consequences are limited such that the need for stringent counter measures, such as evacuation and relocation of the nearby population, can be reasonably excluded. This paper discusses some of the innovative features of the EPR to address severe accident challenges. (author)

  9. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

    2009-09-15

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  10. Severe accident development modeling and evaluation for CANDU

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2009-01-01

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  11. Level 2 PSA methodology and severe accident management

    International Nuclear Information System (INIS)

    1997-01-01

    The objective of the work was to review current Level 2-PSA (Probabilistic Safety Assessment) methodologies and practices and to investigate how Level 2-PSA can support severe accident management programmes, i.e. the development, implementation, training and optimisation of accident management strategies and measures. For the most part, the presented material reflects the state in 1996. Current Level 2 PSA results and methodologies are reviewed and evaluated with respect to plant type specific and generic insights. Approaches and practices for using PSA results in the regulatory context and for supporting severe accident management programmes by input from level 2 PSAs are examined. The work is based on information contained in: PSA procedure guides, PSA review guides and regulatory guides for the use of PSA results in risk informed decision making; plant specific PSAs and PSA related literature exemplifying specific procedures, methods, analytical models, relevant input data and important results, use of computer codes and results of code calculations. The PSAs are evaluated with respect to results and insights. In the conclusion section, the present state of risk informed decision making, in particular in the level 2 domain, is described and substantiated by relevant examples

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

    Directory of Open Access Journals (Sweden)

    Andgren Karin

    2017-01-01

    Full Text Available 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.

  13. Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

    Science.gov (United States)

    Petrosyan, V. G.; Yeghoyan, E. A.; Grigoryan, A. D.; Petrosyan, A. P.; Movsisyan, M. R.

    2018-02-01

    One of the main objectives of severe accident management at a nuclear power plant is to protect the integrity of the containment, for which the most serious threat is possible ignition of the generated hydrogen. There should be a monitoring system providing information support of NPP personnel, ensuring data on the current state of a containment gaseous environment and trends in its composition changes. Monitoring systems' requisite characteristics definition issues are considered by the example of a particular power unit. Major characteristics important for proper information support are discussed. Some features of progression of severe accident scenarios at considered power unit are described and a possible influence of the hydrogen concentration monitoring system performance on the information support reliability in a severe accident is analyzed. The analysis results show that the following technical characteristics of the combustible gas monitoring systems are important for the proper information support of NPP personnel in the event of a severe accident at a nuclear power plant: measured parameters, measuring ranges and errors, update rate, minimum detectable concentration of combustible gas, monitoring reference points, environmental qualification parameters of the system components. For NPP power units with WWER-440/270 (230) type reactors, which have a relatively small containment volume, the update period for measurement results is a critical characteristic of the containment combustible gas monitoring system, and the choice of monitoring reference points should be focused not so much on the definition of places of possible hydrogen pockets but rather on the definition of places of a possible combustible mixture formation. It may be necessary for the above-mentioned power units to include in the emergency operating procedures measures aimed at a timely heat removal reduction from the containment environment if there are signs of a severe accident phase

  14. KAPP-3 and 4 containment pressure following postulated severe accident along with SAMG implementation

    International Nuclear Information System (INIS)

    Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, Nalini; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    Containment is an ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following severe accident i.e. Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet header with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are assumed to be added to calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. After moderator and calandria vault water have completely boiled off, rapid hydrogen generation would take place due to oxidation of pressure tubes and calandria tubes. In such accident scenario, the core is severely damaged. It will also lead to release of a large quantity of radio nuclides to containment atmosphere. To arrest the progression of accident, which can result in Severe Core damage and large amount of hydrogen production, which could leads to containment failure due to hydrogen deflagration or detonation, application of Severe Accident Management Guidelines (SAMG) has been studied. SAMG involve addition of water to calandria and calandria vault. It would result the boiling of the added water and consequent pressurization of containment. This paper presents the analysis for pressure-temperature of KAPP-3 and 4 containment following the postulated accident along with the application of Severe Accident Management Guidelines (SAMG). SAMG initiated action helps in arresting the progression of core

  15. Development of the severe accident management guidance module for the SATS training simulator

    International Nuclear Information System (INIS)

    Kim, K. R.; Park, S. H.; Kim, D. H.

    2004-01-01

    Recently KAERI has developed severe accident management guidance to establish Korea standard severe accident management system. On the other hand PC-based severe accident training simulator SATS has been developed, which uses MELCOR computing code as the simulation engine. SATS graphically displays and simulates the severe accident progression with interactive user inputs. The control capability of SATS makes a severe accident training course more interesting and effective. In this paper the development and functions of HyperKAMG module are explained. Furthermore easiness and effectiveness of the HyperKAMG-SATS system in severe accident management are described

  16. Design measures for prevention and mitigation of severe accidents at advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-06-01

    Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

  17. Maximum forseeable accident analysis made by a sodium leak on the BN-800 primary circuit and the more constraining accident development scenario

    International Nuclear Information System (INIS)

    Ivanenko, V.N.; Zybin, V.A.

    1988-01-01

    In this paper the different ways of development for the BN-800 maximum credible accident in case of loss and fire of primary sodium are examined. The more constraining scenario is presented. During the scenario analysis the accidental release of radioactive materials in the environment has been studied. These releases are below the authorized values [fr

  18. Application of high-order uncertainty for severe accident management

    International Nuclear Information System (INIS)

    Yu, Donghan; Ha, Jaejoo

    1998-01-01

    The use of probability distribution to represent uncertainty about point-valued probabilities has been a controversial subject. Probability theorists have argued that it is inherently meaningless to be uncertain about a probability since this appears to violate the subjectivists' assumption that individual can develop unique and precise probability judgments. However, many others have found the concept of uncertainty about the probability to be both intuitively appealing and potentially useful. Especially, high-order uncertainty, i.e., the uncertainty about the probability, can be potentially relevant to decision-making when expert's judgment is needed under very uncertain data and imprecise knowledge and where the phenomena and events are frequently complicated and ill-defined. This paper presents two approaches for evaluating the uncertainties inherent in accident management strategies: 'a fuzzy probability' and 'an interval-valued subjective probability'. At first, this analysis considers accident management as a decision problem (i.e., 'applying a strategy' vs. 'do nothing') and uses an influence diagram. Then, the analysis applies two approaches above to evaluate imprecise node probabilities in the influence diagram. For the propagation of subjective probabilities, the analysis uses the Monte-Carlo simulation. In case of fuzzy probabilities, the fuzzy logic is applied to propagate them. We believe that these approaches can allow us to understand uncertainties associated with severe accident management strategy since they offer not only information similar to the classical approach using point-estimate values but also additional information regarding the impact from imprecise input data

  19. Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

    International Nuclear Information System (INIS)

    Khalil, Y.F.

    1993-01-01

    Generic Letter 88-20, Supplement No. 1 (Ref. 1), issued by the Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits for nuclear power reactor facilities to perform Individual Plant Examinations (IPE) of their plant(s) for severe accident vulnerabilities and to submit the results to the Commission. This paper summarizes the major Front-End (Level-1 PRA) and Back-End (Level-2 PRA) insights gained from the Millstone Unit 1 (MP-1) IPE study. No major plant vulnerabilities have been identified from a Front-End perspective. The Back-End analysis, however, has identified two potential containment vulnerabilities during postulated events that progress beyond the Design Basis Accidents (DBAs), namely, (1) MP-1 is dominated by early source term releases that would occur within a six-hour time frame from time of accident initiation, or reactor trip, and (2) MP-1 containment is somewhat vulnerable to leak-type failure through the drywell head. As a result of the second finding, a recommendation currently under evaluation, has been made to increase the drywell head bolt's preload from 54 Kips to resist the containment design pressure value (62 psig)

  20. 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.

  1. 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.

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

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    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

  3. Analyses to demonstrate the structural performance of the CASTOR KN12 in hypothetical accident drop accident scenarios

    International Nuclear Information System (INIS)

    Diersch, R.; Weiss, M.; Tso, C.F.; Chung, S.H.; Lee, H.Y.

    2004-01-01

    CASTORc ircledR KN-12 is a new cask design by GNB for KHNP-NETEC for dry and wet transportation of up to twelve spent PWR fuel assemblies in Korea. It received its transport license from the Korean Competent Authority KINS in July 2002 and is now in use in South Korea. It has been designed to satisfy the regulatory requirements of the 10 CFR 71 and the IAEA ST-1 for Type B(U)F packages. Its structural performance was demonstrated against the load cases and boundary conditions as defined in 10 CFR 71 and NRC's Regulatory Guide 7.8, and further explained in NUREG 1617. This included normal conditions of transport load cases - including Hot Environment, Cold Environment, Increased External Pressure (140MPa), Minimum External Pressure (24.5kPa), Vibration and shock, and 0.3m free drop - and the hypothetical accident conditions load cases - including the 9m Free Drop, Puncture, Thermal Fire Accident, 200m Water Immersion and 1.5 x MNOP Internal Pressure. Structural performance were demonstrated by analysis, including state-of-the-art finite element (FE) simulation, and confirmed by tests using a 1/3-scale model. Test results were also used to verify the numerical tool and the methods used in the analyses. All the structural analyses including validation against drop tests were carried out by Arup, and testing were carried out by KAERI. This paper concentrates on the analysis carried out to demonstrate performance in the hypothetical accident 9m free drop scenarios, and results from a small selection of them

  4. Analyses to demonstrate the structural performance of the CASTOR KN12 in hypothetical accident drop accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Diersch, R.; Weiss, M. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany); Tso, C.F. [Arup (United Kingdom); Chung, S.H.; Lee, H.Y. [KHNP-NETEC (Korea)

    2004-07-01

    CASTORc{sup ircledR} KN-12 is a new cask design by GNB for KHNP-NETEC for dry and wet transportation of up to twelve spent PWR fuel assemblies in Korea. It received its transport license from the Korean Competent Authority KINS in July 2002 and is now in use in South Korea. It has been designed to satisfy the regulatory requirements of the 10 CFR 71 and the IAEA ST-1 for Type B(U)F packages. Its structural performance was demonstrated against the load cases and boundary conditions as defined in 10 CFR 71 and NRC's Regulatory Guide 7.8, and further explained in NUREG 1617. This included normal conditions of transport load cases - including Hot Environment, Cold Environment, Increased External Pressure (140MPa), Minimum External Pressure (24.5kPa), Vibration and shock, and 0.3m free drop - and the hypothetical accident conditions load cases - including the 9m Free Drop, Puncture, Thermal Fire Accident, 200m Water Immersion and 1.5 x MNOP Internal Pressure. Structural performance were demonstrated by analysis, including state-of-the-art finite element (FE) simulation, and confirmed by tests using a 1/3-scale model. Test results were also used to verify the numerical tool and the methods used in the analyses. All the structural analyses including validation against drop tests were carried out by Arup, and testing were carried out by KAERI. This paper concentrates on the analysis carried out to demonstrate performance in the hypothetical accident 9m free drop scenarios, and results from a small selection of them.

  5. 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.

  6. A physical tool for severe accident mitigation studies

    Energy Technology Data Exchange (ETDEWEB)

    Marie, N., E-mail: nathalie.marie@cea.fr [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Bachrata, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France); Barjot, F. [EDF R& D, SINETICS, F-93141 Clamart (France); Marrel, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Gossé, S. [CEA, DEN, DPC, F-91191 Gif Sur Yvette (France); Bertrand, F. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France)

    2016-12-01

    Highlights: • Physical tool for mitigation studies devoted to SFR safety. • Physical models to describe the material discharge from core. • Comparison to SIMMER III results. • Studies for ASTRID safety assessment and support to core design. - Abstract: Within the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R&D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic

  7. Estimation of water pollution by domestic in-land nuclear power plant under severe accident

    International Nuclear Information System (INIS)

    Li Hong; Fang Sheng; Fang Dong

    2013-01-01

    In-land nuclear power plant sites of China are usually located in densely populated area and are close to large surface water. This paper proposed scenarios and corresponding calculation models for water contamination caused by radioactive plume release after a severe accident. The models were applied to an imaginary lake (reservoir)-adjacent site in the south of China. The results showed that, the short-time concentration of radioactivity in the lake due to dry and wet deposition and runoff was higher than the generic action levels for foodstuffs in GB 18871-2002, and the public dose resulted was unacceptable. (authors)

  8. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehga, B.R.

    1998-01-01

    This paper describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment integrity to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands in new reactor designs are much stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors

  9. Phebus FP. An international severe accident research programme

    International Nuclear Information System (INIS)

    Hardt, P.; Tattegrain, A.

    1995-01-01

    The main hazard during a hypothetical severe nuclear reactor accident resides in its fission product (FP) inventory. For this reason, the behaviour of FPs has been extensively studied, with the aim of determining the potential source to the environment. The Phebus FP programme proposes a novel, integral approach to this research. After 5 years of construction and of analytical preparation the Phebus FP programme has been supplying a large volume of new experimental data. Their processing by code calculations is presently a major challenge to all partners. The intense collaboration of 25 organizations from 15 countries has proven to be a major asset of Phebus FP. (author). 6 refs., 2 figs

  10. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

    A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

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

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

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

  12. 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

    International Nuclear Information System (INIS)

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

    2017-02-01

    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.

  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. 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.

  15. Multi-phase model development to assess RCIC system capabilities under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kirkland, Karen Vierow [Texas A & M Univ., College Station, TX (United States); Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Beeny, Bradley [Texas A & M Univ., College Station, TX (United States); Luthman, Nicholas [Texas A& M Engineering Experiment Station, College Station, TX (United States); Strater, Zachary [Texas A & M Univ., College Station, TX (United States)

    2017-12-23

    The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that the system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.

  16. Assessment of ICARE/CATHARE V1 Severe Accident Code

    International Nuclear Information System (INIS)

    Chatelard, Patrick; Fleurot, Joelle; Marchand, Olivier; Drai, Patrick

    2006-01-01

    The ICARE/CATHARE code system has been developed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the 'International ICARE/CATHARE Users' Club', under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a 'circuit' experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the stand-alone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version ('Users' Guidelines'). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head. (authors)

  17. Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1991-03-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a pressurized water reactor with a large dry containment. Results from this evaluation include the following: (a) identification of plant conditions that would impact instrument performance and information needs during severe accidents, (b) definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences, and (c) assessment of the availability of plant instrumentation during severe accidents. 16 refs., 3 figs., 4 tabs

  18. Consideration of severe accident issues for the general electric BWR standard plant a status report

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982 the U.S. NRC proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. This paper, GE's submission, discusses the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at TMI, is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  19. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  20. Aircraft accident investigation: the decision-making in initial action scenario.

    Science.gov (United States)

    Barreto, Marcia M; Ribeiro, Selma L O

    2012-01-01

    In the complex aeronautical environment, the efforts in terms of operational safety involve the adoption of proactive and reactive measures. The process of investigation begins right after the occurrence of the aeronautical accident, through the initial action. Thus, it is in the crisis scenario, that the person responsible for the initial action makes decisions and gathers the necessary information for the subsequent phases of the investigation process. Within this scenario, which is a natural environment, researches have shown the fragility of rational models of decision making. The theoretical perspective of naturalistic decision making constitutes a breakthrough in the understanding of decision problems demanded by real world. The proposal of this study was to verify if the initial action, after the occurrence of an accident, and the decision-making strategies, used by the investigators responsible for this activity, are characteristic of the naturalistic decision making theoretical approach. To attend the proposed objective a descriptive research was undertaken with a sample of professionals that work in this activity. The data collected through individual interviews were analyzed and the results demonstrated that the initial action environment, which includes restricted time, dynamic conditions, the presence of multiple actors, stress and insufficient information is characteristic of the naturalistic decision making. They also demonstrated that, when the investigators make their decisions, they use their experience and the mental simulation, intuition, improvisation, metaphors and analogues cases, as strategies, all of them related to the naturalistic approach of decision making, in order to satisfy the needs of the situation and reach the objectives of the initial action in the accident scenario.

  1. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

    2015-01-01

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

  2. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  3. Development of Information Display System for Operator Support in Severe Accident

    International Nuclear Information System (INIS)

    Jeong, Kwang Il; Lee, Joon Ku

    2016-01-01

    When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future

  4. Development of Information Display System for Operator Support in Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Il; Lee, Joon Ku [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future.

  5. Improvement of dose evaluation method for employees at severe accident

    International Nuclear Information System (INIS)

    Onda, Takashi; Yoshida, Yoshitaka; Kudo, Seiichi; Nishimura, Kazuya

    2003-01-01

    It is expected that the selection of access routes for employees who engage in emergency work at a severe accident in a nuclear power plant makes a difference in their radiation dose values. In order to examine how much difference arises in the dose by the selection of the access routes, in the case of a severe accident in a pressurized water reactor plant, we improved the method to obtain the dose for employees and expanded the analyzing system. By the expansion of the system and the improvement of the method, we have realized the followings: (1) in the whole plant area, the dose evaluation is possible, (2) the efficiency of calculation is increased by the reduction of the number of radiation sources, etc, and (3) the function is improved by introduction of the sky shine calculation into the highest floor, etc. The improved system clarifies the followings: (1) the doses change by selected access routes, and this system can give the difference in the doses quantitatively, and (2) in order to suppress the dose, it is effective to choose the most adequate access route for the employees. (author)

  6. Most likely failure location during severe accident conditions

    International Nuclear Information System (INIS)

    Rempe, J.L.; Allison, C.M.

    1991-01-01

    This paper describes preliminary results from which finite element calculation results are used in conjunction with analytical calculation results to predict failure in different LWR vessel designs during a severe accident. Detailed analyses are being performed to investigate the relative likelihood of a BWR vessel and drain line penetration to fail during a wide range of severe accident conditions. Analytically developed failure maps, which were developed in terms of dimensionless groups, are applied to consider geometries and materials occurring in other LWR vessel designs. Preliminary numerical analysis results indicate that if ceramic debris relocates within the BWR drain line to a distance below the lower head, the drain line will reach failure temperatures before the vessel fails. Application of failure maps for these debris conditions to other LWR geometries indicate that in-vessel tube melting will occur in either BWR or PWR vessel designs. Furthermore, if this melt is assumed to fill the entire penetration flow area, the melt is predicted to travel well below the lower head in any of the reference LWR penetrations. However, failure maps suggest the result that ex-vessel tube temperatures exceed the penetration's ultimate strength is specific to the BWR drain line because of its material composition and relatively large effective diameter for melt flow

  7. 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)

  8. Experiments on the lower plenum response during a severe accident

    International Nuclear Information System (INIS)

    Henry, Robert E.; Hammersley, Robert J.; Klopp, George T.; Merilo, Mati

    2004-01-01

    Severe accident evaluations for nuclear reactors consider the response when the core materials have been overheated sufficient to melt and change geometry. One possible consequence of this is that molten core debris could drain into the lower plenum, as occurred in the TMI-2 accident. Given this state, several physical processes need to be analyzed, i.e. the extent of debris particulation and cooling, the potential for thermal attack of lower plenum structures, the thermal transient of the RPV and the potential for external cooling of the RPV lower head. These are important and complex processes, the evaluations of which need to be guided by well founded experiments. To support the development of the MAAP codes, recent experiments have been performed on specific issues such as: 1. the response of lower head penetrations submerged in a high temperature melt, 2. the net steam generation rate when molten debris drains into the lower plenum, 3. the formation of a contact resistance when molten debris drains through water and contacts the RPV wall and 4. the potential for external cooling of the RPV lower head. This paper discusses these experiments and their results. More importantly, it discusses how these are used in formulating models to represent the lower plenum response in the MAAP codes. (author)

  9. Noble gas control room accident filtration system for severe accident conditions (N-CRAFT)

    International Nuclear Information System (INIS)

    Hill, Axel; Stiepani, Cristoph; Drechsler, Michael

    2015-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP either due to containment leakages or due to intentional filtered containment venting. In the latter case aerosols and iodine are retained, however noble gases are not retainable by the FCVS or by conventional air filtration systems like HEPA filters and iodine absorbers. Radioactive noble gases nevertheless dominate the activity release depending on the venting procedure and the weather conditions. To prevent unacceptable contamination of the control room atmosphere by noble gases, AREVA GmbH has developed a noble gas control room accident filtration system (CRAFT) which can supply purified fresh air to the control room without time limitation. The retention process is based on dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. CRAFT allows minimization of the dose rate inside the control room and ensures low radiation exposure to the staff by maintaining the control room environment suitable for prolonged occupancy throughout the duration of the accident. CRAFT consists of a proven modular design either transportable or permanently installed. (author)

  10. ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

    2015-01-01

    ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

  11. Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

    International Nuclear Information System (INIS)

    Cheng, Bo

    2013-01-01

    This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

  12. Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

    International Nuclear Information System (INIS)

    Matsuki, Yoshio; Muramatsu, Ken

    2008-05-01

    The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

  13. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  14. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  15. Noble gas control room accident filtration system for severe accident conditions N-CRAFT. System design

    International Nuclear Information System (INIS)

    Hill, Axel

    2014-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP. This can either be due to leakages of the containment or due to a filtered containment venting in order to ensure the overall integrity of the containment. During the containment venting process aerosols and iodine can be retained by the FCVS which prevents long term ground contamination. Noble gases are not retainable by the FCVS. From this it follows that a large amount of radioactive noble gases (e.g. xenon, krypton) might be present in the nearby environment of the plant dominating the activity release, depending on the venting procedure and the weather conditions. Accident management measures are necessary in case of severe accidents and the prolonged stay of staff inside the main control room (MCR) or emergency response center (ERC) is essential. Therefore, the in leakage and contamination of the MRC and ERC with airborne activity has to be prevented. The radiation exposure of the crises team needs to be minimized. The entrance of noble gases cannot be sufficiently prevented by the conventional air filtration systems such as HEPA filters and iodine absorbers. With the objective to prevent an unacceptable contamination of the MCR/ERC atmosphere by noble gases AREVA GmbH has developed a noble gas retention system. The noble gas control room accident filtration system CRAFT is designed for this case and provides supply of fresh air to the MCR/ERC without time limitation. The retention process of the system is based on the dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. These cycles ensure a periodic load and flushing of the delay lines retaining the noble gases from entering the MCR. CRAFT allows a minimization of the dose rate inside MCR/ERC and ensures a low radiation exposure to the staff on shift maintaining

  16. The once-through mode of steam generator reflux condensation in loss of coolant accident scenarios

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.; Suckow, D.

    2009-01-01

    The once-through mode of steam generator reflux condensation in the presence of noncondensable gases and/or aerosols for LOCA scenarios is introduced. This phenomenon is planned to be investigated at Paul Scherrer Institute in the ARTIST/RFLX experimental program. The plausible accident scenarios associated with the once-through reflux condensation are analyzed with MELCOR to study the safety significance and the boundary conditions of this phenomenon. This work presents the recent PSI experimental and analytical work on reflux condensation: the progress of modification to the ARTIST test facility for the purpose to study reflux condensation, and the analytical model for the once-through reflux condensation in the presence of noncondensable gas using the heat and mass transfer analogy approach. Future experimental and analytical work on reflux condensation is also outlined. (author)

  17. Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

    International Nuclear Information System (INIS)

    Suh, K.Y.; Henry, R.E.

    1996-01-01

    For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

  18. 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

  19. Implementation of hydrogen mitigation techniques during severe accidents in nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    Severe accidents in water-cooled reactors are low-probability events as the Emergency Core Cooling System (ECCS) has been designed and specific accident management measures have been implemented to prevent severe accidents from occurring. Should it not be possible to prevent a severe